72-hour self-sufficiency criterion
72-hour self-sufficiency criterion shall means that the system to which the criterion is applied must be able to perform its function for a minimum of 72 hours so that for the first 24 hours no material replenishments (such as filling the water or fuel tank of the system) are needed, and for the following 48 hours provisions and material reserves exist at the plant site to arrange the necessary material replenishments for the system.
- AICC AICC, Adiabatic Isochoric Complete Combustion, shall refer to a conservative estimate of the pressurisation caused by a hydrogen burn (not a dynamic load, however). Theburn is assumed to be adiabatic (no heat is transferred to the structures), isochoric (volume remains unchanged) and complete (all available hydrogen burns up).
- ATEX ATEX, atmosphères explosibles, shall refer to a potentially explosive atmosphere; the abbreviation Ex means explosive; an Ex space is an explosive space; an Ex component is a component or protection system used in an explosive space.
- Acceptance standard Acceptance standards shall, in the context of in-service inspections, refer to flaw indication acceptance standards presented in ASME Code XI, or to other flaw indication acceptance evaluation methods approved by STUK containing threshold values generally valid for a certain type of components, or parts thereof, not considering the actual stresses present in the item in question.
- Access control Access control shall refer to the access management and guidance of individuals, vehicles and goods by using technical and administrative systems to control access rights of various levels, for example.
Accident shall refer to postulated accidents, design extension conditions and severe accidents. (Nuclear Energy Decree 161/1988)
- Accident modelling method Accident modelling methods, in risk informed planning and assessment of fire protection, are used to collect the results of fire hazard analyses on a case-by-case basis and ensure the adequacy of the nuclear facility’s defence in depth. The methods are used to assess the significance of any fire protection impairments for fire safety of the nuclear facility. • A cause-effect diagram can be used to look for the possible consequences of the selected fire events. • By applying the failure tree and event tree methods, it is possible to define critical events and sequences of events, and assess their significance with regard to the adequacy of the defence in depth and core damage frequency (CDF) of the plant. • Fault and effect analyses and consequence analyses (fire and explosion analyses, dispersion analyses) can be used to assess the sufficiency of the structural and functional layout solutions and other fire protection solutions of the buildings at the plant.
- Accreditation Accreditation shall refer to third-party attestation related to conformity assessment body conveying formal demonstration of its competence to carry out specific conformity assessment tasks. (SFS-EN ISO/IEC 17000, 2005).
- Accreditation body Accreditation body shall refer to a authoritative body that performs accreditation (SFS-EN ISO/IEC 17000, 2005).
- Active failure Active failure shall refer to failure mechanisms other than passive failure mechanisms (such as malfunctions).
- Active fire protection Active fire protection supplements passive fire protection related to the facility’s layout design, fire compartmentation and fire-proof structures. Active fire protection includesfire detection systems and fire extinguishing systems, smoke extraction systems, emergency lighting and operative fire fighting.
- Ageing Ageing shall refer to the potential physical degrading and obsolescence of SSCs at a nuclear facility.
- Ageing management Ageing management shall refer to assuring the operability of SSCs throughout the service life of a nuclear facility. It shall also refer to assuring the conformance with adoptable requirements and current level of technological development.
- Ageing management programme Ageing management programme shall refer to the functions and duties defined by the licensee, pursuant to which the licensee implements the ageing management of a nuclear facility.
- Aggregate Aggregate shall refer to the granular, mineral constituent of concrete that forms concrete when joined together by cement paste.
Air conditioning systems
Air conditioning systems shall refer to systems designed to manage the purity, temperature, humidity and movement of indoor air by treating supply air or circulating air.
- Analysis method Analysis method shall refer to a calculation formula, mathematical modelling method, computer software or other defined workflow that has been proportioned to the difficulty and uncertainties of the task in questionfor the purpose of identifying and describing information and dependencies that affect safety in order to assess the fulfilment of the approval criteria set .
Annual dose shall refer to the sum of the effective dose arising from external radiation within the period of one year, and of the committed effective dose from the intake of radioactive substances within the same period of time. (Nuclear Energy Decree 161/1988)
- Annual limit on intake (ALI) Annual limit on intake (ALI) shall refer to a radionuclide specific maximum value for activity that may enter the body without the annual limit of the effective dose being exceeded. When more than one radionuclides enter the body, the annual limit of the effective dose is not exceeded when the sum of the activities entering the body from all radionuclides divided by the annual limits on intake of the said nuclides does not exceed one.
Anticipated operational occurrence
Anticipated operational occurrence shall refer to such a deviation from normal operation that can be expected to occur once or several times during any period of a hundred operating years. (Nuclear Energy Decree 161/1988)
- Application software Application software shall refer to component software created for each location of use in order to achieve the desired functionality at the location of use. User can usually view or modify application software.
Appropriate certification shall refer to the certification of a quality system based on auditing in which the accreditation of the certification body has been done against the requirements of standard EN ISO/IEC 17021 and the accreditation is covered by the Multilateral Agreements (MLA) or Mutual Recognition Arrangements (MRA) entered into by FINAS.
- Arc Arc shall refer to a physical phenomenon created when the electrical current between two electrodes is discharged through normally non-conducting material, such as air. In the event of an arc, air, which normally insulates electricity becomes conductive due to smoke, for example.
- Artificial defect Artificial defect shall refer to an intentional defect created in the test piece for the purpose of qualification that differs from an actual defect in terms of physical properties. It is usually a groove, indentation or something similar manufactured by means of machining.
- Attenuation function Attenuation function shall refer to a function presenting the acceleration, speed or displacement of the ground motions caused by an earthquake of a certain magnitude as a function of the distance between earthquake's centre and the point of observation, and the frequency of oscillations. The attenuation function can be presented separately for longitudinal and transversal waves.
Auditing shall refer to a systematic, independent and documented process to objectively evaluate the audit evidence obtained to determine the extent to which the agreed auditing criteria are met. (SFS-EN ISO 9000:2015)
Authorised inspection body (AIO)
Authorised inspection body shall refer to an independent inspection organisation approved by the Radiation and Nuclear Safety Authority under Section 60 a of the Nuclear Energy Act to carry out inspections of the pressure equipment, steel and concrete structures and mechanical components of nuclear facilities in the capacity of an agency performing public administrative duties. (Nuclear Energy Decree 161/1988, in Finnish). Authorised inspection body and authorised inspection organisation have same meaning in YVL Guides.
- Automatic fire detection system Automatic fire detection system shall refer to a system that automatically and immediately indicates and locates a starting fire. A fire detection system also provides notification of any failures compromising its functional reliability.
- Auxiliary material Auxiliary welding materials shall refer to, for example, shielding gases and fluxes used in welding.
- Auxiliary system Auxiliary system shall refer to a system required to actuate, control, cool or operate a system executing a safety function, or otherwise maintain the conditions required by the operational prerequisites of the safety function.
Barrier shall refer to an engineered or natural structure or material used for achieving safety functions. A barrier may also be an aggregate formed by different structures and materials.
Baseline configuration shall refer to a configuration of a product, formally established at a specific point in time, which serves as reference for further activities. (ISO 10007)
- Best available techniques Best Available Techniques (BAT) shall refer to methods of production and treatment that are as efficient and advanced as possible and technologically and economically feasible, as well as methods of designing, constructing, maintenance and operation with which the pollution of the environment caused by activities can be prevented or most efficiently reduced.
- Best estimate method Best estimate method shall refer to a method of preparing a safety analysis where the physical modelling of any phenomenon studied is as realistic as possible, and the initial assumptions for the analysis are realistically selected.
- Blind trial Blind trial shall refer to a practical trial conducted under an inspection system to be qualified, witnessed by a qualification body and in which inspectors have not been given advance information about the number, size, orientation or location of defects which the test piece inspected may contain.
- Blow-off valve Blow-off valve shall refer to a valve type used to adjust or limit the pressure of a system or pressure vessel during disturbances.
Boiler shall refer to an assembly intended for the production of steam or the boiling of a liquid other than water to a temperature exceeding 100°C, including at least one piece of heated pressure equipment with the danger of overheating.
Break preclusion (BP)
Break preclusion (BP) shall refer to the principle of using advanced technical and organisational procedures in piping construction, operations and maintenance in such a way that the LBB principle will be implemented and this may be credited for the nuclear power plant design and its provision against the consequences of a complete, instantaneous pipe break.
- Brittle failure Brittle fracture shall refer to a rapid crack growth in a metal structure under tensile stress and with no substantial permanent deformation which, using the energy released, may proceed inside the structure and develop into a complete break.
- Built-to-order product Built-to-order product shall refer to a product designed and manufactured for a special application as single pieces or in small manufacturing batches.
- CE marking CE marking shall refer to the only label that indicates that a building product complies with the declared performance levels and the applicable requirements of the European Union’s harmonised legislation.
- Calibration of a radiation meter Calibration of a radiation meter shall refer to a procedure whereby known kinds of radiation (radiation types and radiation energies) are used to determine the connection between the indication of a meter and the actual value of the radiation variable.
- Catch all material balance area Catch all material balance area shall refer to a material balance area as referred to in Commission Regulation No. 302/2005 that has been approved as part of the EU-wide catch all material balance area (Commission Regulation No. 302/2005, Annex I–G presents the criteria for candidate members of the catch all material balance area).
Certification body shall refer to a recognised third-party organisation under Article 24 of the Pressure Equipment Directive 2014/68/EU.
Nuclear waste other than spent nuclear fuel may, regardless of its radioactive nature, be reused, recycled, recovered and disposed of in accordance with the provisions of the Waste Act (646/2011) if the amount of radioactive substances within it does not exceed the clearance level provided by the virtue of section 7 q, subsection 1, paragraph 28. (Nuclear Energy Act 990/1987 27 c §)
- Collective dose Collective dose
- Commission agreement Commission agreement shall refer to a written agreement containing the information referred to in Section 8 of the Private Security Services Act (282/2002) and Section 10 of the Government Decree on Private Security Services.
- Commissioning Commissioning shall refer to the measures to verify the appropriateness of the licensee's organisation as well as the planned operation and safe use of the plant and its systems, structures and components.
- Commissioning inspection Commissioning inspection shall refer to an inspection that ensures the operability of safety class 2 and 3 concrete structures, steel and concrete structures, prestressed concrete structures, or steel and composite structures and buildings and that is performed before the nuclear operation of the facility begins.
- Commissioning testing Commissioning testing shall refer to tests that ensure that the plant and its systems, structures and components function as planned. Commissioning testing is part of the commissioning. For a nuclear power plant, commissioning testing can be divided into the following main parts, for example: system tests, fuel loading andpre-criticality tests, making the reactor critical, lower power tests, and power tests.
- Common cause failure Common cause failure shall refer to a failure of two or more structures, systems and components due to the same single event or cause.
- Common cause failure in PRA In PRA common cause failure shall refer to a scenario where the following three conditions are met: 1) Two or more individual systems, components or structures have failed or degraded so that they are no longer operable when necessary. 2) The failures overlap temporally so that the fulfilment of the success criteria in a random demand is uncertain. 3) The failures are caused by a common cause or mechanism, but they are not consequential failures.
- Competence Competence shall refer to a person’s knowledge and skills, suitability for his or her position, attitude towards and understanding of the safety significance of his or her work, and an ability to apply such competence to duties of safety significance.
Competence development shall refer to a set of planned, target-oriented and structured actions designed to assess, maintain and develop individual and organisational competences.
- Competence management Competence management shall refer to set of actions designed to ensure the acquisition, development, and maintenance of knowledge and skills, including systematic assessment, development, and utilisation of these actionsin the management of an organisation.
- Complete, instantaneous break Complete, instantaneous break shall refer to the breaking of a pipe or another type of break that results in a large leak due to structural damage, degradation of material properties, or overload.
- Component life cycle Component life cycle shall refer to the various stages of a component from design to production, operation, maintenance and decommissioning.
- Composite structure Composite structure shall refer to a structural entity formed by a concrete and steel structure where the interoperation of the concrete and the steel components have a substantial role in relation to the load-bearing capacity, leak-tightness and fire protection properties of the structural entity.
Conceptual plan of in-service inspections
Conceptual plan of in-service inspections shall refer to a conceptual plan required under Section 35 of the Nuclear Energy Decree (161/1988). It refers to a document describing the in-service inspections performed using non-destructive inspection methods throughout the entire life cycle of a nuclear facility, from design to decommissioning. The conceptual plan of in-service inspections contains a preliminary description of the risk-informed selection processes of inspection items, the selection criteria for inspection intervals, the inspection systems and their qualifications, and the methods used to report and assess inspection results and flaw indications.
- Concrete Concrete shall refer to material that has been fabricated by mixing cement, coarse and fine aggregate and water, and potentially admixtures and additives, and the properties of material develop as the cement hardens (hydrates) with the help of water.
- Concrete cover Concrete cover shall refer to the thickness of the concrete layer protecting the reinforcement.
- Concrete element Concrete element shall refer to a concrete structure that has been cast and cured outside of its final location (either factory-built or fabricated at the construction site).
- Concrete structure Concrete structure shall refer to concrete, reinforced concrete and prestressed concrete structures.
- Condition monitoring Condition monitoring shall refer to the determining of the operability of a SSC.
Conditional core damage probability
Conditional core damage probability (CCDP) shall refer to the probability of core damage as a result of an initiating event.
Conditional large release probability
Conditional large release probability (CLRP) shall refer to the probability of a large release due to an initiating event.
- Conformity assessment Conformity assessment shall refer to the demonstration that the requirements (SFS-EN ISO/IEC 17000, 2005) relating to a product, process, system, person or body are fulfilled.
- Conformity assessment body Conformity assessment body shall refer to a body that implements conformity assessment services. (SFS-EN ISO/IEC 17000, 2005)
- Consequence of failure Consequence of failure shall refer to the conditional core damage probability caused by a pipe leak. The consequences of failure are divided into consequence categories.
- Consequential failure Consequential failure shall refer to a failure caused by a failure of another system, component or structure or by an internal or external event at the facility
- Conservative analysis method Conservative analysis method shall refer to a manner of preparing a safety analysis that considers the uncertainties related to the calculation models and initial assumptions so that, with a high level of certainty, the consequences of the event analysed would be milder than the analysis results.
- Construction inspection Construction inspection shall refer to the verification of the requirements prescribed in the product’s construction plan.
- Construction of a nuclear facility Construction of a nuclear facility (construction project) shall refer to the measures that the holder of a construction licence has taken in order to construct a nuclear facility that conforms to the requirements set until the end of the commissioning of the facility.
- Construction plan Construction plan shall refer to the design documentation compiled for the purpose of pre-inspection conducted by STUK or an authorised inspection body.
Construction plans for nuclear fuel, control rods and dummy elements
Construction plans for nuclear fuel, control rods and dummy elements shall refer to the written documentation defining the detailed requirements for the structure and manufacturing of fuel, control rods and dummy elements, and their inspections taking place during manufacturing.
- Construction project for a nuclear facility Construction project for a nuclear facility shall refer to the construction of a new nuclear facility or modifications at an operating nuclear facility (including commissioning).
- Construction project organisation Construction project organisation shall refer to the organisation required to construct a nuclear facility meeting the requirements set.
- Containment function Containment systemshall refer to the containment (structure) and its systems that are designed to isolate the containment, remove heat from inside the containment, and control radioactive substances and combustible gases in accident scenarios.
- Contamination Contamination refers to undesirable radioactive substances on surfaces (surface activity), or within solids, liquids or gases (also in the human body).
Contributory factors shall refer to events or conditions that, together with the other causes, increase the probability of an event but will not cause the event alone.
- Control of manufacturing Control of manufacturing shall refer to a process to monitor the progress of manufacture to ensure that a product or delivery follows all designs and plans.
- Controlled area Controlled area shall refer to a working area in which specific radiation protection procedures shall be followed and to where access is controlled.
Controlled state shall refer to a state where a reactor has been shut down and the removal of its decay heat has been secured. (STUK Y/1/2018)
Controlled state following a severe reactor accident
Controlled state following a severe reactor accident shall refer to a state where the removal of decay heat from the reactor core debris and the containment has been secured, the temperature of the reactor core debris is stable or decreasing, the reactor core debris is in a form that poses no risk of re-criticality, and no significant volumes of fission products are any longer being released from the reactor core debris. (STUK Y/1/2018)
- Core damage frequency Core damage frequency (CDF) shall refer to the expectation value of a core damage event per unit of time.
Correction shall refer to a action that is performed in order toeliminate observed detected nonconformity.
Corrective actions shall refer to actions intended to eliminate the cause of nonconformities and prevent the recurrence of nonconformities. (SFS-EN ISO 9001:2015)
- Corrosion Corrosion shall refer to a physical and chemical reaction between metal and its environment that introduces changes to the metal's properties and may lead to a significant reduction in the functionality of the metal, its environment, or the technical system of which they are part.
- Critical defect size Critical defect size shall refer to a size of an assumed crack-like defect in pressure equipment which, if exceeded, would, according to the applicable fracture mechanical criterion, cause a risk of fast fracture in a load scenario where the combination of the stress states and material properties is the least favourable.
Criticality shall refer to a state where the output and loss of neutrons, created in nuclear fission and maintaining a chain reaction, are in equilibrium so that a steady chain reaction continues. (STUK Y/1/2018)
Criticality accident shall refer to an accident caused by an uncontrolled chain reaction of nuclear fissions. (STUK Y/1/2018)
- Criticality safety index Criticality safety index shall refer to a figure related to the design of a fissile material package that limits the number of packages in one transport unit.
Damping ratio shall refer to the ratio of the actual damping coefficient (the ratio of the viscous damping force to velocity) for a single-degree-of-freedom oscillator to the critical damping coefficient (the maximum value of the damping coefficient at which periodically attenuating oscillation is possible). The damping ratio is usually expressed as a percentage.
- Dangerous object Dangerous object shall refersuch an object, copyof an object, or substance that may endanger or can be used to endanger the safety or security of a nuclear facility or persons witjhin the nuclear facility, or the safety of persons participating in the treatment and transport of nuclear use items or nuclear waste. (Government Decree 734/2008)
Decommissioning shall refer to the dismantling of a permanently closed nuclear facility so that no special measures are required at the facility site due to radioactive materials originating from the dismantled facility. (Nuclear Energy Act 990/1987)
- Decommissioning waste Decommissioning waste shall refer to the low and intermediate level waste arising from the dismantling of a nuclear facility.
- Decontamination Decontamination shall refer to cleaning radioactive substances from components, structures or rooms.
- Defence in depth approach to fire protection The aim of the defence in depth approach to fire protection is to prevent the breakout of fires, detect and extinguish fires quickly, prevent the development and spreading of fires, and limit their effects so that the safety functions can be performed reliably irrespective of the effects.
- Degradation mechanism Degradation mechanism shall refer to a phenomenon or process that may cause degradation in a pressure-bearing structure.
Derived air concentration (DAC)
Derived air concentration shall refer to a radionuclide-specific maximum value for the average airborne activity concentration, under which 2,000 hours of work may be carried out annually without exceeding the dose limits.
Design bases shall refer to all requirements, definitions and bases for normal operational conditions and accidents that pertain to the design and operation of a plant, system and component. (Nuclear Energy Decree, 161/1988)
- Design basis earthquake Design basis earthquake shall refer to facility site ground motion used as the basis for the nuclear facility’s design. The design basis earthquake shall be so defined that in the current geological conditions the anticipated frequency of occurrence of stronger ground motions is not more often than once in a hundred thousand years (1×10-5/a) at median confidence level. A design basis earthquakes are represented using peak ground acceleration and ground response spectra.
- Design basis fire Design basis fire shall refer to the worst possible fire situation the probability of which during the design period is not negligible. It is taken into account in the design of fire protection systems, such as fire compartmentation, the fire water mains, and the fire extinguishing systems. A design basis fire must always be determined if the size of a fire load contained by a fire compartment and involved in combustion is assumed to be lower that the fire load of the entire fire compartment. The design basis fire must be justified using hazard, failure and impact analyses.
- Design basis scenario Design basis scenario shall refer to the reactor's normal operation, anticipated operational occurrences, postulated accidents, and design extension conditions.
Design basis threat
Design basis threat shall refer to a threat of unlawful action used as the basis for the planning and assessment of the nuclear security arrangements for which the licensee is responsible. (161/1988)
Design extension condition
Design extension condition shall refer to:
a. an accident where an anticipated operational occurrence or class 1 postulated accident involves a common cause failure in a system required to execute a safety function;
b. an accident caused by a combination of failures identified as significant on the basis of a probabilistic risk assessment; or
c. an accident caused by a rare external event and which the facility is required to withstand without severe fuel failure.
(Nuclear Energy Decree 161/1988)
- Design load Design loads, to be determined in accordance with the applicable standard, comprise design pressure, design temperature and other mechanical design loads which in combination with design pressure produce the highestprimary stresses in normal operating conditions.
- Design organisation Design organisation shall refer to any organisation involved in design activities, including any design modifications.
- Destructive testing Destructive testing shall refer to inspections performed in order to determine the mechanical characteristics of a itemand which destroy or transform the geometry of the piece inspected.
- Dimensioning calculation Dimensioning calculations (strength calculations) shall refer to defining the main dimensions of the structure using mechanical loads, allowable stresses and deformations; dimensioning calculations are also used to design a structure so that it is appropriate and meets all requirements.
- Direct causes Direct causes shall refer to immediate events or conditions that cause an event.
- Direct operational activities Direct operational activities shall refer to all activities that directly concern the systems, structures and components of a nuclear power plant.
- Direct visual testing Direct visual testing shall refer to inspections performed without the use of additional instruments (with the exception of light sources, mirrors and magnifying glasses).
Disposal facility shall refer to an entierety comprising therooms for the disposal of the waste packages (emplacement rooms) and the adjoining underground and above-ground auxiliary facilities. (Nuclear Energy Decree 161/1988)
Disposal site shall refer to the location of the disposal facility and, after disposal has been completed, the area entered in the real estate register in accordance with Section 85 of the Nuclear Energy Decree, and the ground and bedrock under it.
Diversity principle shall refer to the backing up of functions through systems or components having different operating principles or differing from each other in some other manner, with all systems or components able to implement a function separately. (STUK Y/1/2018)
- Division of inspection responsibilities Division of inspection responsibilities shall refer to the division of inspections of mechanical components between STUK, an inspection organisation and a third party.
- Dose commitment Dose commitment
Dose constraint shall refer to a constraint on the individual radiation dose of a person other than a patient arising from ionizing radiation during a specific period of time, used to optimize radiation protection in radiation practices. (Radiation Act 859/2018)
Dose limit shall refer to the radiation dose arising from ionizing radiation which may not be exceeded during a specific period of time. (Radiation Act 859/2018)
Dose Registry shall refer to a file into which the dose information and identification information of the employees engaged in radiation work is saved.
Dual-use item shall, in the context of Finnish legislation, refer to a product, technology, service or other commodity that can, in addition to its civilian use or application, be used to develop or manufacture weapons of mass destruction or missile systems that can be used to guide them to their targets.
- Dynamic analysis Dynamic analysis shall refer to determining the time-dependent behaviour (vibrations) and stresses of a component or structure under an impact-type, seismic or cyclic loading. In particular, the analysis focuses on the resonance risk induced by the excitation of natural oscillations and the strengthening of stress in proportion to the stresses caused by an equivalent static load.
- Dynamic analysis (hoisting equipment) Dynamic analysis shall, in the context of Guide YVL E.11, refer to the definition of the eigenvalues, acceleration, displacement and hydraulic stresses of components and structures. The values calculated are used as input data for the strength analysis and dimensioning as well as to ascertain the durability of various parts under anticipated loading conditions.
- EIA procedure EIA procedure shall refer to an environmental impact assessment procedure whereby the environmental impact of certain projects is investigated and assessed, and the views of authorities and those whose circumstances or interests may be affected by the project, as well as the communities and foundations whose field of activity may be affected by the project, are heard.
Effective dose shall refer to the weighted sum of the equivalent doses in tissues and organs exposed to radiation, where equivalent dose denotes the product of the mean energy imparted by radiation to tissue or to an organ, per unit mass, and a weighting factor specified for the radiation. Effective dose is presented as a formula in the Government Decree on Ionising Radiation (1034/2018).
- Electrical equipment Electrical equipment shall refer to equipment used in the production, transmission and transformation of electrical power and the protection of the grid. Electrical equipment includes accumulators, transformers, distribution centres, power distribution network protective relays, motors, frequency converters and electromechanical components. If a nuclear facility uses distributed instrumentation and control systems, with I&C functions distributed between various pieces of electrical equipment, such as protective relays and frequency converters, the requirements of I&C equipment shall also be taken into account in handling these electrical equipment.
- Emergency arrangements Emergency arrangements shall refer to advance preparation for accidents or events impairing safety at the nuclear facility or in its site area or other places or vehicles where nuclear energy is used. (Nuclear Energy Act 990/1987)
Emergency exposure situation
Emergency exposure situation means a situation in which the consequences of a radiation safety incident require or may require special measures to limit or reduce the radiation exposure of persons participating in the emergency work or protective measures or the exposure of members of the public. (Radiation Act 859/2018)
- Emergency lighting Emergency lighting shall refer to back-up lighting and emergency exit lighting, including lighting for exit routes, open areas and hazardous working areas; emergency lighting functions simultaneously with conventional lighting and independent of it.
Emergency planning zone
Emergency planning zone shall refer to an area extending to a distance of approximately 20 kilometres from the nuclear power plant and for which authorities shall draft an external rescue plan referred to in Section 48(1)(1) of the Rescue Act (379/2011). (STUK Y/2/2018)
- Emergency power supply Emergency power supply shall refer to a system consisting of a diesel engine or a gas turbine and a generator and all systems, structures and components required for its start-up and operation. Whenever a specific part of an emergency power supply is discussed in this Guide, it is mentioned separately.
Emergency situation shall refer to an accident or event during which the nuclear power plant’s safety has deteriorated or is in the danger of deteriorating or requires enhanced preparedness to act in order to ensure plant safety; emergency situations are classified on the basis of their severity and controllability as follows:
- an alert is a situation where the safety level of a nuclear power plant needs to be ensured in an exceptional situation.
- a site area emergency is a situation during which the nuclear power plant’s safety deteriorates or is in the danger of deteriorating significantly.
- a general emergency is a situation during which there is danger of radioactive substance releases that may require protective measures in the vicinity of the nuclear power plant.
(STUK Regulation Y/2/2018)
- Encapsulation Encapsulation shall refer to the operations related to spent nuclear fuel to be enclosed in a final disposal canister at an encapsulation plant.
- Encapsulation plant Encapsulation plant shall refer to a nuclear facility that is used to encapsulate spent nuclear fuel for final disposal.
- Equipment for use of force Equipment for use of force shall primarily refer to a firearm referred to in the Firearms Act (1/1998), a gas spray referred to in the Firearms Act, handcuffs, a baton not more than 70 cm in length, and a telescopic baton, the latter referring to a baton over 45 cm and not more than 70 cm in length when opened and that can be collapsed for transport.
- Essential parameters Essential parameters shall, in the context of qualification, refer to the influential parameters whose value changes may actually influence an inspection in a manner where the goals of the inspection can no longer be achieved. They include input information parameters, inspection procedure parameters, and inspection equipment parameters.
- Eurocodes Eurocodes shall refer to the pan-European design standards for load-bearing structures published by the European Committee for Standardization.
- European Technical Assessment (ETA) European Technical Assesment (ETA) shall refer to an approval that may be granted for building products for which no harmonised product standard exists. ETA is a voluntary technical approval resulting in CE marking.
- Event analysis Event analysis shall refer to the identification of faults and the factors resulting in faults.
- Event investigation Event investigation shall refer to investigating the progress, causes and consequences of an event, identifying the factors resulting in the event, and estimating the measures required to prevent the occurrence of similar events. Systematic and documented methods must be used in the investigation.
- Event notification Event notification shall refer to a written account provided by a security guard/employee containing the information listed in Section 17 of the Private Security Services Act and Section 11 of the Government Decree on Private Security Services.
- Excepted package Excepted package shall refer to a package that may contain limited amounts of radioactive materials, instruments, or articles.These limited amounts are defined in the transport regulations for dangerous goods; such as with regard to road transport in Decree 369/2011, Annex A, Table 18.104.22.168.4.1.2.
- Exclusive use Exclusive use shall mean the sole use, by a single consignor, of a conveyance or of a large freight container, in respect of which all initial, intermediate and final loading and unloading and shipment are carried out in accordance with the directions of the consignor or consignee.
- Execution (construction) Execution of construction shall refer to all functions required for the physical creation of a building, including the acquisitions made for the site, the manufacture, transport, storage and installation of building materials and products, and inspection activities and documentation.
- Execution class (construction) Execution class shall refer to a clas¬sified collection of requirements set for execution, which may apply to the entire construction work, a single assembly or a detail of a single assembly (SFS-EN 1090-2, SFS-EN 13670).
- Execution specification (construction) Execution specification shall refer to a pres¬entation of the information and requirements that are necessary for the execution of steel structures, including the necessary in¬formation and requirements concerning steel structural products and steel structure com¬ponents. Execution specification for a concrete structure shall refer to a classified collec¬tion of requirements set for execution, which may apply to the entire construction work or a single component.
- Existing software Existing software shall refer to software that has been manufactured before the life cycle of the present project, such as the software for a serially manufactured pressure transducer or the system platform software of programmable logic.
Expected evolution shall refer to evolution where the disposal system performs as planned or where one or more long-term safety functions are assumed to degrade.
- Expert panel Expert panel, in the context of in-service inspections, shall refer to a special workgroup consisting of experts of various fields of nuclear facility technology that evaluates risk-informed methods.
- Explosion An explosion can be a fast chemical reaction (detonation) that releases a large amount of thermal and kinetic energy. A vapour explosion occurs when liquefied gas is vaporised extremely rapidly, causing a mechanical explosion.
External events shall refer to exceptional situations or incidents occurring in the vicinity of a nuclear facility that could have a detrimental effect on the safety or operation of the plant.
Facility shall refer to a nuclear facility or any other location where nuclear materials are customarily used. Facility also refers to a nuclear facility or a location where the annual material flow or use will exceed one effective kilogram. (IAEA Safeguards Glossary, 2001 Edition, International Nuclear Verification Series No. 3)
- Factory test (hoisting equipment) Factory test shall refer to the functional testing performed on a hoisting device unit, including load tests, which is extensive enough to ascertain the proper functioning of the product and which takes the restraints resulting from the physical size and factory conditions of the product into consideration.
Failure criterion (N+1)
(N+1) failure criterion shall mean the same as the single failure criterion. Single failure criterion (N+1) shall mean that it must be possible to perform a safety function even if any single component designed for the function fails.
Failure criterion (N+2)
(N+2) failure criterion shall mean that the most important safety functions necessary to bring the plant to a controlled state and to maintain it must be ensured in postulated accidents even if any individual component of a system providing the safety function is inoperable and even if any other component of a system providing the same safety function or of a supporting system necessary for its operation is simultaneously inoperable due to the necessity for its repair, maintenance or testing.
- Failure potential Failure potential shall refer to the proneness of a structure to degradation mechanisms. Failure potentials are divided into failure categories.
- Fatigue Fatigue shall refer to damage propagating in local structural discontinuities as a result of cyclic mechanical or thermal loads, which is evaluated in a stress analysis by comparing the peak stresses calculated for these locations to the fatigue curve of the applicable standard.
- Fault, failure Fault or failure, in the context of Guide YVL A.8, shall mean that a SSC no longer meets the operability requirements.
- Final disposal canister Final disposal canister shall refer to a hermetic, corrosion resistant and mechanically strong container where spent nuclear fuel is enclosed.
- Final suitability assessment Final suitability analysis shall refer to the licensee’s assessment to demonstrate (validate) that a component meets its rated values. This can be achieved by means of type approval and testing, quality control procedures, and operating experience.
- Finite element method Finite element method (FEM) shall refer to a mathematical modelling method of physical phenomena in structures and other continua, in which the domain to be analysed and the distribution of the variables governingthe phenomena are described numerically, by using a meshof finite elements.
- Fire Fire shall refer to a chemical reaction that produces heat, smoke, or both. The inherent characteristics of a fire include uncontrolled development and spreading.
- Fire compartment Fire compartment shall refer to a section of a building from which the propagation of fires has been prevented by means of separating, fire resistant structural elements.
- Fire hazard analysis • A structural fire hazard analysis (FHA) is conducted to ensure the sufficient fulfilment of the defence in depth approach to fire protection by means of validated deterministic methods. • A functional fire hazard analysis (FFHA, FHFA) is conducted to ensure the sufficient fulfilment of the defence in depth principle to fire protection in the functional layout solutions for the implementation of the plant’s safety functions. • Analyses of computational fluid dynamics (CFD) simulate the gas flows in fires, as well as fire development and heat transfer. • The finite element method (FEM) is used to simulate the conduction of fire-induced heat in solid matter and the development of structural responses as a result of the temperature.
- Fire load Fire load shall refer to the total thermal energy released by the materials in a space (fire compartment) when the combustible material in that space is fully burned. Fire load density shall refer to the total thermal energy of the fire load per unit of floor area.
- Fire protection system Fire protection systems shall refer to fire detection systems, fire-fighting water pumping stations , the fire water mains, extinguishing systems, smoke extraction systems, and emergency lighting.
- Fire protection threat Fire protection threat shall refer to a situations that can result in the outbreak of a fire or an explosion resulting from a fire load.
- Fissile material Fissile material shall, in accordance with the regulations for the transport of dangerous goods (such as 369/2011, Appendix A, Chapter 22.214.171.124.3), refer to fissile nuclides, including uranium-233, uranium-235, plutonium-239, plutonium-241 and all materials that contain one of these. However, unirradiated natural uranium, depleted uranium, ornatural uranium only irradiated in a thermal reactor shallnot be considered fissile material under the transport regulations.
- Flaw indication Flaw indication shall, in the context of in-service inspections, refer to evidence of a flaw obtained using a non-destructive inspection method.
Floor response spectrum
The floor response spectrum describes the maximum vibrations of single-degree-of-freedom oscillators with various natural frequencies and a particular damping ratio positioned in a certain area of a building. The calculation of the floor response spectrum involves an analysis, based on the dynamic behaviour of the frame structure, of the transfer of the vibration from an earthquake or other external source to the part of the building under examination where it strains systems, structures, and components. Depending on the design basis, the floor response spectrum is an acceleration, velocity or displacement spectrum or an energy response spectrum in accordance with their combination.
- Fracture prevention Break preclusion (BP) shall refer to the principle of using advanced technical and organisational procedures in piping construction, operations and maintenance in such a way that the LBB principle will be implemented and this may be credited for the nuclear power plant design and its provision against the consequences of a complete, instantaneous pipe break.
- Fracture support Whip restraint shall refer to a steel structure constructed to provide protection against a breaks inhigh-energy pipes by preventing impacts caused by the break on components and structures important to safety, and by limiting the hydrodynamic load caused on the internal structures of the same system by a sudden leakage flow.
- Fragility curve (seismic) Fragility curve (seismic) shall refer to a curve that describes the probability of component or structure failure as a function of ground acceleration.
- Fresh concrete Fresh concrete shall refer to fully mixed concrete that is still in a state where it may be compacted using a method of choice.
- Fuel failure Fuel failure shall refer to a scenario where a fuel rod loses its integrity.
- Fuel item Fuel item shall, in the context of Guide YVL D.1, refer to a fuel assembly, fuel rod or corresponding batch of nuclear material; unit shall refer to any batch of nuclear material.
- Function critical to safety Function critical to safety shall refer to a work stage or task with a bearing on nuclear safety or radiation safety and, owing to its nature, calls for special competence and personal characteristics, such as an ability to concentrate, accuracy and stress tolerance.
Function important to safety
Functions important to safety shall refer to functions where the individual performing the function has a direct or indirect effect on the safety of the nuclear facility at the various stages of its life cycle.
Functional isolation shall refer to the isolation of systems from one another so that the operation or failure of one system does not adversely affect another system; functional isolation also covers electrical isolation and isolation of the processing of information between systems. (STUK Y/1/2018)
- Functional test (hoisting equipment) Functional tests shall, in the context of Guide YVL E.11, refer to the functional tests and testing conducted in the latter stage of the commissioning inspection in accordance with an approved commissioning test programme at the location of use of the hoisting device unit.
General equipment requirement specification
General equipment requirement specification shall refer to a document that includes general equipment group-specific design and quality control requirements for safety classes 1, 2 and 3. When an equipment is procured, the requirements set out in this document will be completed with site-specific requirements.
Genericinvestigation shall refer to an analysis or investigation using information concerning several similar events.
- Geometrical indication Geometrical indication shall refer to an indication of the geometrical or metallurgical structure of the component that is obtained by using non-destructive testing methods.
- Graded approach Graded approach related to nuclear security shall refer to a principle according to which, the specification, planning and implementation of nuclear security takes into account the applicable threat assessment, the properties of nuclear materials, and the potential consequences of unlawful action directed at nuclear materials.
- Ground response spectrum Ground response spectrum shall refer toa method of presentation that describes the maximum vibrations of single-degree-of-freedom oscillators assumed to be anchored in site bedrock at various natural frequencies and using a particular damping ratio.
- HCLPF (High confidence of low probability of failure) capacity HCLPF (High confidence of low probability of failure) capacity shall refer to a ground acceleration value that results in a component or structure failure probability of 5% at a confidence level of 95%.
- Harmonised product standard Harmonised product standard shall refer to a product standard drawn up by the European Committee for Standardization (CEN) that results in CE marking and of which an announcement has been published in the Official Journal of the European Union. For each product group, it defines the properties that are determined for the products, the manufacture quality control requirements, and the information to be reported in the CE marking.
- Hazard curve Hazard curve shall refer to a curve that describes the frequency of exceedance (probability per year, for example) of a particular parameter value. A hazard curve can be presented for multiple statistical confidence levels.
High level waste
High level waste shall refer to waste that, because of its high level of activity, requires highly effective radiation protection arrangements when processed and usually also cooling. The activity concentration of such waste is usually more than 10 GBq/kg.
- High strength concrete High strength concrete shall refer to concrete that has a compression strength class higher than C50/60 pursuant to standard EN 1992-1-1.
- Hoisting accessory Load-lifting attachment shall refer to a device or structure used to assist in hoisting or transfer.
- Hoisting device Load-handling device shall refer to a part of the hoisting device unit to which the load to be moved is fixed directly or using a hoisting accessory.
- Hoisting device unit Hoisting device unit shall refer to hoisting and transfer devices, including the components that are part of their mechanical structure and rail track, instrumentation and control and electrical systems related to their operation, and other fixed auxiliary systems (hydraulic, pneumatic etc.),
- Hold point Hold point shall refer to an inspection for which advance invitations have been sent to the parties defined in the inspection plan and whose supervision is a condition for proceeding with the work unless the parties have given written permission to proceed without their presence.
- Hydrodynamic design Hydrodynamic design shall refer to hydrodynamic dimensioning in normal use as defined in the process requirements; hydrodynamic design also refers to transient and accident analyses when the geometry, dimensions and material of the structure are known.
I&C (Instrumentation and Control) equipment
Instrumentation and control (I&C) equipment shall refer to a piece of equipment such as a measuring device, display unit, electronic card or I&C relay that is part of an instrumentation and control system.
I&C (Instrumentation and Control) platform
Instrumentation and control (I&C) platform shall refer to a product that, by the use of pre-developed tools and software and electronic modules, enables the design, implementation and maintaining of an application-specific I&C system.
INES scale (International Nuclear and Radiological Event Scale) shall refer to the IAEA's international classification scale of nuclear facility events and radiological events used to estimate and illustrate the radiation and nuclear safety significance of events when communicating on them. The INES scale is not applied to events which have no influence on radiation or nuclear safety.
- IRS system IRS system (Joint IAEA/NEA International Reporting System for Operating Experience) shall refer to the operating experience reporting system maintained by the IAEA and in use in its member states. The system is used to report events and experiences related to operation and decommissioning, the causes of which and lessons learned have significance on a global scale.
- Ignition Ignition shall refer to a chemical reaction that can result in a fire. An open flame, sparks, the heating of electrical systems or cables, hot surfaces or static electricity can be sources that cause ignition.
- In-service inspection plan for an inspection interval In-service inspection programme for an inspection interval shall refer to an inspection programme with a scope of one inspection interval, such as ten years. The programme for each inspection interval presents the inspections selected for performance during the inspection interval in question, and the inspection procedures and other documents to which changes have been introduced after the previous interval.
- In-service inspection plans for the operation period In-service inspection plan for the operation period shall refer to an inspection plan presenting the inspections performed during an operation period. The inspections are performed during the refuelling and service outages that determine the length of the operation period.
- Independent assessment of management system Independent assessment of a management system shall refer to an assessment performed by an internal or external party independent of the management, implementation and development of the item assessed. In general, independence can be demonstrated by indicating that the assessor is not responsible for the function being assessed.
- Indication Indication shall, in the context of in-service inspections, refer to flaw indications and geometrical indications.
- Indirect visual examination Indirect visual testing shall refer to an inspection performed using e.g a boroscope, endoscope, fibre optics, binoculars or a TV camera. An automated control and data recording system, manipulator, or robot may be used to assist in the inspection.
- Influential parameter Influential parameters shall, in the context of non-destructive in-service inspections, refer to the parameters that are significant to a specific inspection and may affect the results of the inspection.
- Information security management system An information security management system shall refer to a part of the nuclear facility's general management system that is created and implemented, used, supervised, reviewed, maintained, and continuously improved. The information security management system comprises the structure of the organisation, the information security management policy, planning activities, responsibilities, procedures, methods, processes, and resources.
- Initiating event Initiating event shall refer to an identified event that leads to anticipated operational occurrences or accidents.
- Inspection Inspection shall refer to the examination of components or structures and related designs and processes as well as the verification of their conformity to requirements in terms of the requirements presented in STUK’s decisions, the YVL Guides and the design bases.
Tarkastuslaitos tarkoittaa laitosta, joka suorittaa tarkastuksia tuotteen, prosessin, palvelun tai asennuksen tai niiden suunnittelun tutkimiseksi ja sen vaatimustenmukaisuuden todentamiseksi. (SFS-EN ISO/IEC 17020:2012)
Inspection organisation approved by STUK
Inspection organisation approved by STUK shall refer to an inspection organisation that has been approved in accordance with Guide YVL E.12.
- Inspection procedure Inspection procedure shall, in the context of in-service inspections, refer to a written description and definition of a method of performing an inspection in a particular test situation, specifying all essential parameters and defining the factors to be observed when applying an inspection technique in line with established standards, procedures and regulations.
- Inspection reliability Inspection reliability shall, in the context of in-service inspections, refer to the degree which an inspection system achieves in the detection, characterisation and sizing of defects at an acceptable false call rate.
- Inspection system Inspection system shall refer to all those elements of non-destructive testing that may influence the quality and outcome of an inspection, such as inspection equipment and their software, inspection procedures and personnel. Qualification of an inspection system shall, in the context of in-service inspections, refer to the systematic evaluation of a system, using all methods that are needed to provide reliable confirmation that an inspection system is capable of preforming as required under real inspection conditions. An inspection system is qualified by demonstrating that it reliably detects, characterises and/or determines defects endangering structural integrity and nuclear safety so that the inspection goals set forth in the input information are met.
- Installation construction plan (buildings and structures) Installation construction plan shall (for buildings and structures), refer to a construction plan describing how a steel structure, steel component of a composite structure, or a concrete element or a precast concrete product, including supports, connects to other structures.
- Installation construction plan (lifting and handling equipment) Installation construction plan shall, in the context of Guide YVL E.11, refer to a construction plan describing how a hoisting device unit connects to a nuclear facility, including the electrical and I&C systems.
- Installation construction plan (pressure equipment) Installation construction plan shall, in the context of Guide YVL E.3, refer to a construction plan describing the connecting of the pressure equipment to the rest of the system and including supports.
Instrumentation and control (I&C) system
Instrumentation and control (I&C) system shall refer to a system that controls, adjusts or observes the operation of a nuclear facility or components thereof. Control may consist of automatic operation or relaying commands given by humans. The results of the observation may be processed by the I&C system or relayed to the operators of the facility.
Intermediate level waste
Intermediate level waste shall refer to nuclear waste that, because of its high level of activity, requires effective radiation protection arrangements when processed. The activity concentration of such waste is usually 1 MBq/kg – 10 GBq/kg. (STUK Y/4/2018)
- Internal contamination Internal contamination shall refer to radioactive substances inside the body.
Internal events shall refer to events occurring inside a nuclear facility that may have an adverse effect on the safety or operation of the plant.
- Laboratory approved by STUK Laboratory approved by STUK shall refer to a laboratory whose calibration can be used to demonstrate the traceability of measurements into the international measurement system. Traceability shall, in this context, refer to the connection between a measurement result on the meter and the measurement standards through an uninterrupted chain of comparisons where uncertainties have been reported for each comparison.
- Large release frequency Large release frequency (LRF) shall refer to the expectation value of the frequency of a large release of radioactive substances per unit of time.
“Leadership” shall refer to a manager using their skills and capabilities to direct the operation of employees and groups and have an impact on their commitment, values and behaviour. “Management” shall refer to authorised actions to guide the organisation and ensure work performance.
Leak Before Break (LBB)
The leak before break (LBB) shall refer to a principle where piping does not have identified failure mechanisms creating the possibility of complete break, and faults that are not detected during inspections may at most create a small, local leak the detection of which allows the plant to be brought into a state where there is no risk of complete break.
- Level of qualification Qualification level shall refer to a reference level of inspection reliability set as an objective for an inspection system which will be qualified. The qualification level defined in the input information is dependent on the nuclear-safety risk significance of the failure of the structure and the role of the inspection in the reduction in the probability of failure. The failure potential of the structure and the consequences of the failure affect the nuclear safety risk.
Licensee shall refer to the holder of a licence entitling to the use of nuclear energy. (Nuclear Energy Act 990/1987)
- Licensee under a waste management obligation Licensee under a waste management obligation shall refer to a licensee whose operations generate or have generated nuclear waste. (Nuclear Energy Act 990/1987)
- Licensee’s in-house inspection organisation Licensee's in-house inspection organisation shall refer to the licensee's separate inspection unit, the position of which is arranged in compliance with the type B requirements of ISO/IEC EN 17020, the operations of which meet the specific requirements laid down by STUK, and which STUK has approved to carry out inspection tasks pertaining to the pressure equipment, steel and concrete structures and mechanical components of a nuclear facility in the form of in-house control by the licensee.
- Licensee’s personnel Licensee’s personnel shall refer to the personnel necessary for the use of nuclear energy who are employed by the licensee.
- Load analysis Loading analysis shall refer to the computational analysis, covering the entire life cycle, of the mechanical and thermal loads (service loads) to which a component is subjected in the operational conditions and accidents used as the facility's design bases over the course of its entire life cycle, when the procedures, specifications and analyses concerning operation, required functions and sequences of events are taken into account.
- Load grouping Load grouping shall refer to applying the acceptance limits presented in the standard applied to stress analysis, graded by the severity of the load and the safety factors required, to the specified service loads so that the frequency of occurrence of the load, the post-load opportunities for inspection and repair, and the integrity and operability requirements set for the pressure equipment in the scenario in question are taken into consideration.
- Loading test (hoisting equipment) Loading test shall refer to a load test or overload test performed on a hoisting device in accordance with a chosen standard where the machinery and structures of the hoisting device are loaded.
- Location requirements Location requirements shall refer to the requirements set on a component by its location of use at a nuclear facility. The location of use at a facility and as part of the facility’s other systems sets requirements on the device in terms of the tolerance of environmental conditions, functionality, performance, and tolerance of potential accident conditions.
Long-lived waste shall refer to nuclear waste the calculated activity concentration of which after 500 years is above 100 MBq/kg in a disposed waste package, or above an average value of 10 MBq/kg in waste placed in one emplacement room.
Long-term safety shall refer to the safety of disposal after the closure of the disposal facility, taking account of radiation exposure on humans and the environment.
Long-term safety functions
Long-term safety functions shall refer to functions achieved by the characteristics or processes of engineered and natural barriers that are intended to isolate the nuclear waste from the bedrock and the biosphere or to impede the migration of radionuclides.
Loss of coolability of the nuclear fuel
Loss of coolability of the nuclear fuel shall refer to a scenario where nuclear fuel loses its coolable shape due to fuel failure or a deformation that exceeds the design basis, or where a flow cooling fuel rods is blocked due to impurities in the fuel assembly.
Low energy equipment
Low energy equipment shall refer to Safety Class 2 equipment with a design pressure of up to 20 bar(g) and a design temperature of up to 120 °C and to which the design, dimensioning and quality-control requirements of a corresponding equipment from Safety Class 3 can be applied with technical justifications without having a risk to lose the operability of the equipment.
Low level waste
Low level waste shall refer to nuclear waste that, because of its low level of activity, can be processed without any special radiation protection arrangements. The activity concentration of such waste is usually not more than 1 MBq/kg. (STUK Y/4/2018)
- Low power test Low power tests shall refer to tests performed at low power after nuclear fuel has been loaded into the reactor, with the objective of ensuring that the reactor core behaves as expected, that the plant and its systems, the reactivity control systems in particular, operate as intended, and that it is safe to proceed to higher power levels. During the tests, the reactor power shall remain at the lowest possible level where the necessary parameters can be reliably measured.
- Maintenance Maintenance shall refer to the planned service of SSC to reduce the probability of failure in advance, or the overhaul or repair of a SSC undertaken on the basis of observed needs.
- Making the reactor critical Making the reactor critical shall refer to the following actions:• in a pressurised water reactor, the the boron dilution from primary coolant is begun; • in a boiling water reactor the, withdrawal of the first control rod from the reactor is begun.
Management shall also refer to a person or group orienting and guiding the organisation. Management includes top and middle management.
Management system shall refer to a system that is used to establish policy and objectives and to achieve those objectives.
- Manufacturer Manufacturer shall refer to an individual or organisation responsible for the design, manufacture, testing, inspection and installation of equipment or sets of assemblies. A manufacturer may subcontract one or more of the said tasks under its responsibility.
- Manufacturer approved by STUK Manufacturer approved by STUK shall refer to a manufacturer approved under Guide YVL E.3.
- Manufacturing (building products) Manufacturing shall, in the context of building products, refer to all the measures that are required to produce and deliver a steel assembly. Depending on the case, this may include procurement, pre-fabrication and assembly, welding, mechanical attachment, transport, surface treatment, inspection, and documentation.
- Material balance area Material balance area shall refer to an area, as referred to in Commission Regulation No. 302/2005, for which the European Commission has granted a material balance area code after the information required in the Regulation has been submitted.
- Material manufacturer Material manufacturer shall refer to an individual or organisation producing material, in basic product form, used in the manufacturing of a component or structure.
Mechanical components shall refer to, for example, internals of reactor pressure vessel, control rod drives, pumps, motors, filters, valve actuators, cranes, load lifting attachments, fuel handling equipment, final disposal canisters, fans, and air ducts.
- Mechanical load Mechanical load shall refer to the pressure, external forces and moments which satisfy the laws of equilibrium between external and internal forces and moments.
- Minimal cut set Minimal cut set shall, in a level 1 PRA, refer to the smallest combination of an initiating event, failures or errors which can result in core damage.
Minimum system performance
Minimum system performance can be determined by making the following assumptions:
- Consider the consequential effects of the initiating event (component failure, for example).
- Furthermore, select the failure combination that is most detrimental to the functionality of the system in accordance with the failure criterion presented in requirement 442 of Guide YVL B.1. The single failure with the highest reactivity effect is also assumed to occur in the reactor scram system.
- Determine the performance parameters for each functioning component, which conform to the acceptance limits of components in periodic tests.
- Modelling of inspections Modelling of inspection shall refer to the use of mathematical prediction models of non-destructive inspections in order to quantitatively assess the performance of inspection systems as part of the technical justification.
- Modification Modification shall refer to introducing changes to a system, structure or component so that it no longer corresponds to previous specifications.
- Modification (emergency power supply) Modification shall refer to the alteration of an emergency power supply in such a manner that it no longer corresponds to the earlier designs. The acquisition of a spare part that is significant in terms of operability is also considered a modification whenever the structure or material of the spare part differs from the original.
- Modular structure Modular structure shall refer to an assembly consisting of several structural components and equipment parts. Components of the structure may include steel components, piping and piping supports, valves, cable trays, steel platforms and reinforcement steel. As the load-bearing structure of a modular structure actually the steel components delivered to the construction site as pre-fabricated parts, or a composite structure consisting of these steel assemblies and concrete.
Natural release barrier
- Non-destructive testing Non-destructive testing shall refer to inspections that do not essentially alter the geometry and size of the item inspected.
- Non-linear structure analysis Non-linear structural analysis shall refer to solving a partial differential equation group using the finite element method (FEM), for example. The non-linearity of a structure may be geometrical by nature, due to the behaviour of the material, or related to the boundary conditions. A geometrical non-linearity is created when the loads cause large displacements in relation to the dimensions of the structure; in this case, the distribution of stress in the non-deformed and deformed structure will vary greatly. The non-linearity of a material means that the constitutive equations of the material are not linear.
- Non-nuclear pressure equipment Non-nuclear pressure equipment shall refer to pressure equipment classified in class EYT under Guide YVL B.2.
Normal operating conditions
Normal operating conditions shall refer to the planned operation of a nuclear facility according to the operating procedures. Normal operating conditions also include testing, plant start-up and shutdown, maintenance and the replacement of nuclear fuel. (STUK Y/1/2018)
YVL Guides also use the term normal operation, which means the same as normal operating conditions.
- Normal power supply systems Normal power supply systems shall refer to power supply systems whose operation is not secured by safety-classified auxiliary power supply systems located within the plant site.
Notified body shall refer to a notified body as referred to in Article 12 of the Pressure Equipment Directive (2014/68/EU).
Nuclear facility shall refer to the facilities used for the generation of nuclear energy, including research reactors, facilities for the large-scale disposal of nuclear waste, and facilities for the large-scale production, use, processing or storage of nuclear material and nuclear waste. However, nuclear facility shall not refer to:
a) mines or ore processing plants intended for the production of uranium or thorium, or premises and locations including their precincts where nuclear wastes from such facilities are stored or deposited for final disposal; or
b) facilities and premises that have been permanently closed and where nuclear waste has been disposed in a manner approved as permanent by the Radiation and Nuclear Safety Authority; or
c) premises or parts of a nuclear facility that have been decommissioned in a manner approved by the Radiation and Nuclear Safety Authority. (Nuclear Energy Act 990/1987)
- Nuclear facility waste Operational waste shall refer to the low and intermediate level waste arising from the operation of nuclear facilities.
- Nuclear facility’s pressure equipment Nuclear facility’s pressure equipment shall refer to nuclear and non-nuclear pressure equipment at a nuclear facility.
Nuclear information shall refer to:
• software that has been specially designed or modified for the development, production or use of the goods referred to in Appendix A to the Nuclear Energy Decree
• technology which means special information, in written or other form, necessary for the development, production or use of the goods referred to in Appendix A to the Nuclear Energy Decree, and that is not generally available or not related to basic scientific research. This information takes the form of technical data or technical assistance. Technical data may take forms such as blueprints, plans, diagrams, models, formulae, tables, engineering designs and specifications, manuals and instructions written or recorded on other media or devices such as disk, tape, read-only memories. Technical assistance may take forms such as instructions, skills, training, working knowledge and consulting services and may involve the transfer of technical data. (Council Regulation (EC) No 428/2009) The Nuclear Energy Act shall apply to nuclear information which may have bearing on nuclear proliferation or which are subject to the obligations of Finland’s international treaties in the nuclear energy sector. However, a licence defined in the Nuclear Energy Act is only needed for nuclear information that are subject to a particular safeguards obligation (Nuclear Energy Act, Section 19 a).
- Nuclear liability Nuclear liability shall refer to the liability of an operator of a nuclear facility for damage incurred to a third party.
Nuclear material shall refer to special fissionable materials and source materials, such as uranium, thorium and plutonium, suited for obtaining nuclear energy. (Nuclear Energy Act 990/1987, Section 3).
- Nuclear material category Nuclear material category shall refer to natural uranium (N), depleted uranium (D), low enriched uranium (below 20%, L), high enriched uranium (20% or above, H), plutonium (P) and thorium (T).
Nuclear power plant
Nuclear power plant shall refer to a nuclear facility for the purpose of electricity or heat production, equipped with a nuclear reactor, or a complex consisting of nuclear power plant units and other related nuclear facilities located at the same plant site.
- Nuclear pressure equipment Nuclear pressure equipment shall refer to pressure equipment classified in safety class 1, 2 or 3 under Guide YVL B.2.
- Nuclear safeguards Nuclear safeguards shall refer to regulatory control preventing the proliferation of nuclear weapons to ensure that the nuclear materials and nuclear energy are used peacefully as defined in international treaties, and to ensure that they or any technology related to them is not used to promote the proliferation of nuclear weapons.
Nuclear use item
Nuclear use item shall refer to nuclear material and the substances, devices, equipment, nuclear information and agreements referred to in Sections 2(1)(5) and 2(2)(1) of the Nuclear Energy Act (990/1987). (Nuclear Energy Decree 161/1988)
Nuclear waste shall refer to a) radioactive waste in the form of spent nuclear fuel or in some other form generated during or as a result of the use of nuclear energy and b) materials, objects and structures which, having become radioactive during or as a result of the use of nuclear energy and having been removed from use, require special measures owing to the danger posed by their radioactivity. (Nuclear Energy Act 990/1987)
Nuclear waste facility
Nuclear waste facility shall refer to a nuclear facility utilised for the encapsulation of spent nuclear fuel or the processing of other nuclear waste for disposal, and to a disposal facility for spent nuclear fuel or other nuclear waste; however, nuclear facility shall not refer to permanently closed facilities where nuclear waste has been placed in a manner approved as permanent by the Radiation and Nuclear Safety Authority. (Nuclear Energy Decree 161/1988)
Obsolescence shall mean that a plant SSC fails to meet the new safety requirements or no longer represents the current technological development in terms of the assurance of safety. The lack of technical support or spare parts is also regarded as a manifestation of the obsolescence of a SSC.
- Open trial Open trial shall, in the context of qualification, refer to a practical trial conducted using the inspection system under qualification which is witnessed by a qualification body and where the inspectors have been given advance knowledge of the defects of the test piece inspected.
- Operability Operability shall refer to the integrity and performance of SSC in conformance with its design bases.
- Operating organisation Operating organisation shall refer to an organisation and personnel required for the safe operation of a nuclear facility.
- Operational activities of a nuclear power plant Operational activities of a nuclear power plant shall refer to the operations subject to an operating licence engaged in by a licensee to operate a nuclear power plant. The operational activities of a nuclear power plant are considered to commence when measures to load nuclear fuel into the reactor are launched.
Operational event shall refer to a failure, flaw or non-conformity in safety functions, systems, components, structures or an organisation’s activities that has a bearing on radiation safety or nuclear safety. Operational events also include emergencies and disturbances as well as events compromising radiation safety. Operational events also include events taking place during the construction phase. Note: Operational event is a wider concept, but in YVL Guides, it shall refer to events in STUK’s controlled area (radiation and nuclear safety).
Operational limits and conditions, OLC
The Operational Limits and Conditions (OLC)) set out the technical and administrative requirements for ensuring the plant’s operation in compliance with the design bases and safety analyses; the requirements for ensuring the operability of systems, structures and components important to safety; and the limitations that must be observed in the event of component failure.
- Operative fire fighting Operative fire fighting comprises, for example, the plant fire brigade as well as the fire fighting measures conducted by the plant personnel and other fire brigades in the nearby areas. It covers the fire fighting equipment of the plant fire brigade, as well as the fire fighting equipment at the plant, such as portable fire extinguishers and fire hydrants. Fire fighting plans, rescue diving instructions, training and instructions on fire fighting as well as fire drills are all encompassed by operative fire fighting.
Operator shall, depending on the context, refer to a licensee, licence applicant or some other user of nuclear energy (use of nuclear energy is defined in Section 2 of the Nuclear Energy Act) who is engaged in, or is planning to engage in, operations falling within the area of application of the Nuclear Energy Act.
- Operator (transport) Operator shall, depending on the context, refer to a licensee, licence applicant or some other user of nuclear energy (Section 2 of the Nuclear Energy Act) who is engaged in, or is planning to engage in, operations falling within the area of application of the Nuclear Energy Act. In the transport of nuclear material or nuclear waste, the user of nuclear energy specified above shall be the consignor, carrier, or consignee as per prior agreement concerning the division and transfer of responsibilities between the consignor, carrier, and consignee and considering the responsibilities for different parties under the transport regulations for dangerous goods.
- Optimisation principle Optimisation principle means that operation is organised such that radiation exposure harmful for health is kept as low as is reasonably achievable.
As referred to in Section 2(1)(6) of the Nuclear Energy Act, ores containing uranium or thorium shall mean minerals in which:
1) the average content of uranium is higher than 1 kilogram per metric tonne; or
2) the average content of thorium is higher than 30 kilograms per metric tonne, except for monazite, or higher than 100 kilograms per metric tonne for monazite.
(Nuclear Energy Decree 161/1988, Sections 2 and 3(2)(3)).
- Organisational competence Organisational competence shall refer to the performance level of an organisation and related factors (such as structures, resources, processes, networks).
- Overhaul Overhaul shall refer to the elimination of the non-conformities or shortcomings observed in the structure or performance of a SSC as the SSC still fulfils the requirements set for its operability.
- Package Package, in connection with radioactive materials, shall refer to packaging and its radioactive contents in the form it has been submitted for transport.
- Packaging Packaging shall refer to parts necessary to completely enclose the radioactive contents.
- Packed waste Packed waste shall refer to waste that has been conditioned into its final form for storage and/or disposal.
Parametric studies shall, in the context of qualification, refer to experimental laboratory studies to establish the impact of various separate essential parameters. The testing organisation conducts them to support the technical justification without witness by a qualification body.
- Particular safeguards obligation Particular safeguards obligation shall refer to an obligation arising from a bilateral nuclear agreement concluded by Finland or the European Union with some other state or a group of states outside the EU which covers the possession, manufacture, production, transfer, handling, use, storage, transport, export or import of nuclear material, nuclear waste, ore or some other material, component, equipment or technology as referred to in Section 8(1) (Nuclear Energy Decree, 161/1988).
Pass/fail criterion for qualification during in-service inspection
Pass/fail criteria for in-service inspection qualification shall refer to the qualification assessment criteria relating to the number of defects detected in a test piece, the accuracy of determining defect size and position, and other factors that are reported in an inspection qualification and that determine the acceptability of the inspection system.
- Passive failure Passive failure shall refer to a mode of failure that can be treated as an operability deficiency (such as a total or partial lack of a device or operability).
Passive fire protection
Passive fire protection shall refer to structural fire protection arrangements, such as fire compartmentation together with the functional design and layout design of the facility, in order to ensure the facility’s safety during fire situations as far as possible without active fire fighting operations.
- Peak ground acceleration (PGA) Peak ground acceleration (PGA) shall refer to the highest acceleration of ground motion due to an earthquake.
- Peer review group Peer review group shall refer to a team of experts from other facilities who perform an inspection.
Performance target shall refer to a measurable or assessable characteristic of a barrier. The performance target shall include a criterion describing the characteristic which, when met, ensures the performance of a long-term safety function.
- Permanent joint Permanent joint shall refer to a joint which can only be removed using destructive methods.
- Personnel and work shift records Personnel and work shift records shall refer to the bookkeeping records referred to in Section 18 of the Private Security Services Act (282/2002) that are intended to ensure official supervision by authorities.
- Physical degrading Physical degrading shall refer to the degradation of structural or functional characteristics in use or with time as a result of physical, chemical and/or biological mechanisms. Physical degrading may lead to the loss of operability of a SSC.
- Physical reasoning Physical reasoning shall, in the context of qualification, refer to a part of the technical justification that summarises the detailed selection justifications of the qualification approach expressed in qualitative terms. The design of test pieces may be started using physical reasoning.
- Physical security function Physical security function shall refer to the design, installation, repair or modification of structural protection or electronic surveillance systems as well as the planning of other physical protection arrangements (282/2002) and the supervision of these tasks.
- Physical security task Physical security task shall refer to the nuclear security and supervision tasks performed by security staff, as laid down in Sections 7 l and 7 m of the Nuclear Energy Act.
Physical separation shall refer to the separation of systems or components from one another by means of adequate barriers, distance or placement, or combinations thereof. (STUK Y/1/2018)
- Physical separation in connection with fire protection Physical separation in connection with fire protection shall refer to the separation of spaces or components by means of distance, fire resistance, fire-proof structures, local fire insulation, or a combination thereof.
- Pipe routing instructions Pipe routing guideline shall refer to pre-defined procedureon ensuring the acceptability of thelocation of the piping, piping supports and equipmentso that thermal expansions and stresses do not damage piping or cause leaks.
- Piping Piping shall refer to assemblies used to transport gas, liquid, steam and mixtures thereof, where piping components are connected together for integration into a pressure system.
- Piping flexibility analysis Piping flexibility analysis shall refer to determining piping thermal expansion and calculation of reaction forces and moments at piping supports and equipment joints so that thermal expansion and stresses do not damage piping or cause leaks.
- Piping part Piping components shall refer to separate elements such as a pipe or a system of pipes, tubing, fittings, expansion joints, hoses and other pressure retaining components.
- Piping segment Piping segment shall refer to a piping section where a breakage or leak caused by a failure of any point of the piping section has the same consequences and where the degradation mechanisms are the same.
- Piping stress analysis Piping stress analysis shall refer to the stress analysis conducted according to the piping design standard.
- Piping structural element Piping structural element shall refer to a part of a given piping segment, such as a weld joint, straight piping section, pipe bend, fitting, flange connection, or bellows.
- Plant area Plant area shall refer to an area that consists of a fenced area containing buildings related to the operation of the facility, and it shall be located within a zone where movement and sojourn are limited (described in subsection 3.4.2 of Guide YVL A.11).
- Plant fire brigade The plant fire brigade shall refer to a fire brigade maintained by the licensee of a nuclear power plant site. Its purpose is to extinguish fires occurring at the nuclear power plant, as well as the plant site and its immediate vicinity, maintain operative fire fighting preparedness, monitor fire and chemical safety, and provide training to plant personnel on fire protection.
- Plant modification Plant modification (plant modification project) shall refer to a modification to the safety-classified systems of an operating nuclear facility that calls for the reassessment of the design bases and safety requirements of systems, the renewal of the design basis analyses, and considerable equipment acquisitions. Examples of plant modifications include a power increase of the reactor or a modernisation of the protection I&C systems.
Postulated accident shall refer to a deviation from normal operation which is assumed to occur less frequently than once over a span of one hundred operating years, excluding design extension conditions; and which the nuclear facility is required to withstand without sustaining severe fuel failure, even if individual components of systems important to safety are rendered out of operation due to servicing or faults. Postulated accidents are grouped into two classes on the basis of the frequency of their initiating events: a) Class 1 postulated accidents, which can be assumed to occur less frequently than once over a span of one hundred operating years, but at least once over a span of one thousand operating years; b) Class 2 postulated accidents, which can be assumed to occur less frequently than once during any one thousand operating years. (Nuclear Energy Decree 161/1988)
- Power supply systems Power supply system shall refer to systems designed to supply the necessary electrical power to the actuators and instrumentation and control systems of the plant unit.
Practical trials shall, in the context of qualification, refer to the assessment of non-destructive inspection by open and blind trials, witnessed by a qualification body and applying the inspection to test pieces containing defects. Laboratory tests performed as parametric studies that are not witnessed by a qualification body are not practical trials.
- Pre-criticality tests Pre-criticality tests shall refer to tests that can only be performed after nuclear fuel has been loaded, but before the reactor is made critical. These include for example testing the pressure loss of the main circulation flow and control rod movement using actual core geometry.
- Pre-passivation Pre-passivation shall refer to a procedure whereby a thin, protective oxide layer is formed on the surface of a material when metal reacts with its environment. The formation of an oxide layer reduces the rate of corrosion.
Pre-service inspection plan
Pre-service inspection plan shall refer to an inspection plan the purpose of inspections of which is to provide basic comparative data for in-service inspections and obtain data supplementing manufacture and installation quality assurance about the original condition of the inspection items inspected within the in-service inspection scope.
- Precast concrete product Precast concrete product shall refer to a concrete element manufactured under an applicable European product standard.
Precautionary action zone
Precautionary action zone shall refer to an area extending to a distance of approximately five kilometres from the nuclear power plant, where land use restrictions are in force. (STUK Y/2/2018)
- Prefabrication Prefabrication shall refer to any actioncarried out on materials used when they are transformed into components that are completed to be installed and joined in the assembly.
Preliminary suitability analysis
Preliminary suitability analysis shall be used by the licensee to verify that a component is suitable for its intended location of use on the basis of its rated values. The qualification of the component is also inspected and designed. After the preliminary suitability analysis the requirement specification of the component isverified, and the procurement of the component may be started, if necessary.
- Pressure control Pressure control shall refer to pressure regulation, overpressure protection and pressure reduction.
- Pressure control analysis Pressure control analysis shall refer to an analysis used to demonstrate that the pressure control systems meet the design requirements set for them.
- Pressure dimensioning Pressure design shall refer to determining the material thickness required of pressure equipment under pressure by means of calculation formulae presented in the applicable standard.
- Pressure equipment Pressure equipment shall refer to a vessel, piping and other technical assembly, in which overpressure exists, or in which it may develop, as well as the technical assemblies designed to protect pressure equipment, including elements attached to pressure retaining parts such as flanges, nozzles, couplings, supports, lifting lugs etc.
- Pressure equipment construction plan The documentation to be submitted for regulatory consideration before the commencement of the manufacture of pressure equipment. It documents the structure of the pressure equipment, including calculations, manufacturing plans, inspection and testing plans, the suitability of other related components, and a summary of licensee's justification for acceptability.
Pressure equipment dossier
Pressure equipment dossier shall refer to a document defined in Section 69 of the Pressure Equipment Act (1144/2016) containing all the protocols and other essential documentation pertaining to pressure equipment.
Pressure equipment subject to registration
Pressure equipment subject to registration shall refer to pressure equipment to be registered under Section 51 of the Pressure Equipment Act (1144/2016).
- Pressure vessel Pressure vessel shall refer to a housing designed andbuilt to contain fluids under pressure , including direct attachments up to the coupling point connecting it to other equipment. A vessel may contain one or several chambers.
- Pressure vessel stress analysis Pressure vessel stress analysis shall refer to the stress analysis conducted according to the pressure vessel design standard.
- Pressurised auxiliary equipment Pressure accessory shall refer to devices with an operational function and having a pressure-bearing housing, such as valves, pressure regulators, measuring chambers, pressure gauges, glass gauges, filters and expansion joints.
- Prestressed concrete structure Prestressed concrete structure shall refer to a reinforced concrete structure with renforcement that is partially or fully prestressed.
- Prestressed tendon Prestressing tendon shall refer to the basic material used for reinforcement of concrete structures subject to prestressing.
- Prestressing system Prestressing system shall refer to the complex formed by the prestressing tendons used, the installation, prestressing, locking, and protection of the prestressing tendons , and the related equipment and work methods.
Preventive actions shall refer to measures intended to eliminate the cause of potential deviations and prevent the occurrence of deviations. (ISO 9001)
- Primary containment Primary containment shall refer to a pressure-proof and leak-tight building surrounding the reactor and its coolant circuit, the function of which is to protect the reactor and the coolant circuit from external events and prevent the release of radioactive substances into the environment in accidents. When the word ‘containment’ is used in Guide YVL B.6, it refers to the primary containment. The primary containment may be surrounded by a secondary containment. The purpose of the secondary containment is to make possible the recovery and processing of any radioactive substances leaking form the primary containment. For this purpose, the interim space between the primary containment and secondary containment is kept at underpressure. The secondary containment may also provide protection against external events.
- Primary stress Primary stress shall refer to stress generated in a pressure equipment structure that keeps the structure in balance in relation to the external mechanical loads exerted on it.
- Principle of limitation Principle of limitation means that the radiation exposure of an individual shallnot exceed the maximum values confirmed by the Radiation Decree.
- Probabilistic fire risk assessment Probabilistic fire risk assessment shall refer to a quantitative assessment of the fire-induced hazards, probabilities and negative consequences of sequences of events influencing the safety of a nuclear power plant.
- Probabilistic fracture mechanics model Probabilistic fracture mechanics (PFM) model shall refer to a methodology used to determine the failure probabilities of degraded pressure-retaining structures. When using a deterministic reliability model to calculate the failure time of a degraded pressure-retaining structure, the essential initial parameters (such as loads, fracture strength, original failure frequency, fracture growth pattern and so on) of several single events are input into the model using representative probability distributions of the parameters selected. The time-dependent failure probability is determined by calculating the ratio between the number of failure cases of damaged pressure-retaining structures which have occurred by the time of assessment and the number of calculated cases.
Probabilistic risk assessment (PRA)
Probabilistic risk assessment (PRA) shall refer to a quantitative assessments of hazards, probabilities of event sequences and adverse effects influencing the safety of a nuclear power plant. (Nuclear Energy Decree 161/1988)
Process is a set of interrelated or interacting activities which transforms inputs into outputs.
Product shall refer to a result of a process. Examples of products include a nuclear facility, plant modification, system delivery, single component or part thereof, plan or design, processed material, or information product. A service is not a product.
- Production test Production test shall refer to a test whereby a sample corresponding to a joint or coating is welded during welding work in order to ensure that the properties of the product meet the requirements set.
Products specific to a delivery batch
Products specific to a delivery batch shall refer to products (materials, parts, components) that have been allocated to a nuclear fuel delivery batch at the time of their manufacture. Other nuclear fuel products are not considered allocated to a delivery batch.
Project shall refer to a unique process consisting of a set of coordinated and controlled activities with start and finish dates, undertaken to achieve an objective conforming to specific requirements, including the constraints of time, cost and resources. (SFS-EN ISO 9000:2015) Projects include for example construction projects of new nuclear facilities or plant modification projects launched to implement extensive modifications.
Protection I&C systems
Protection automation shall refer to I&C systems that control systems needed to manage postulated accidents and design extension conditions in order to reach a controlled state and maintain it. Protection I&C systems also include I&C systems that control systems needed for accident management during possible I&C common cause failures.
- Pump Pump shall, in the context of Guide YVL E.9, refer to an entity (pump unit) including a hydraulic machine transferring liquid, its drive and any other structures and and parts that are essentially involved in the entity.
- Pump NPSH Pump NPSH (Net Positive Suction Head) shall refer to the difference between the pressure at the intake flange and the steam pressure corresponding to the temperature of the fluid pumped. Pump NPSHa shall refer to an NPSH available at the service place, and the NPSHr shall refer to an NPSH that the pump requires to operate without the adverse effects of cavitation.
Qualification is normally used as a synonym for “validation” in YVL-guides. Qualification shall refer to confirmation, through the provision of objective evidence, that the requirements for a specific intended use or application have been fulfilled.
Qualification shall refer to a demonstrated ability to apply knowledge and skills.(SFS-EN ISO 9000:2015)
- Qualification approach of in-service inspections Qualification approach of in-service inspections shall refer to a combination of qualification measures that is required to achieve the required level of qualification. The level of difficulty and novelty of the inspections and the level of qualification required are considered when defining the approach.
- Qualification body Qualification body shall refer to an independent expert body that plans, conducts, assesses and witnesses qualifications of inspection systems.
- Qualification certificate Qualification certificate shall refer to a document issued under the rules of an inspection qualification system indicating that there exists an adequate level of confidence that the inspection equipment, procedures and personnel or any combination of these are capable, in a specific inspection, of achieving the inspection objectives set.
- Qualification dossier for in-service inspection Qualification dossier of in-service inspection shall refer to a collection of all relevant information concerning the definition and implementation of inspection qualification. The qualification dossier includes information on defects, items inspected, inspection conditions and the inspection procedure. It also covers the qualification procedure, technical justification, and the results of the qualification.
- Qualification input information for in-service inspection Input information for qualification of in-service inspection shall refer to the information, such as the essential parameters describing an item inspected and the objectives of the inspection, that must be available before inspection qualification is started.
- Qualification of inspection system Qualification of an inspection system shall, in the context of in-service inspections, refer to the systematic evaluation of a system, using all methods that are needed to provide reliable confirmation that an inspection system is capable of preforming as required under real inspection conditions. An inspection system is qualified by demonstrating that it reliably detects, characterises and/or determines defects endangering structural integrity and nuclear safety so that the inspection goals set forth in the input information are met.
- Qualification organisation Certification body shall refer to a recognised third-party organisation under Article 13 of the Pressure Equipment Directive 97/23/EC.
- Qualification procedure Qualification procedure shall refer to a systematic sequence of rules describing how a certain non-destructive test performed on a particular structure is to be qualified.
- Qualification system Qualification system shall refer to a system that includes the procedures and administration to perform qualification tasks.
- Quality management Quality management shall refer to all of the coordinated and planned activities performed to ensure that the organisation, component, plant or activity meets the requirements and quality criteria set for it (SFS-EN ISO 9000).
Quality management system
Quality management system shall refer to a management system to direct and control an organisation with regard to quality. (SFS-EN ISO 9000:2015).
- Radiation embrittlement Radiation embrittlement shall refer to the embrittlement of the reactor pressure vessel's steel in the core region due to the microstructural damage and transition temperature increase (radiation shift) caused by neutron radiation.
- Radiation exposure Radiation exposure shall refer to being exposed to radiation.
Radiation safety incident
Radiation safety incident means an event or situation that compromises or may compromise radiation safety and unplanned medical exposure. (Radiation Act 859/2018)
- Radioactive substance Radioactive material in the regulations for the transport of dangerous goods shall refer to a material (including fissile materials, other nuclear materials, or nuclear waste) the activity concentration and total activity of which in the consignment exceed the radionuclide-specific limits defined in the transport regulations.
- Random failure Random failure shall refer to a failure the events of which cannot be anticipated other than by means of statistical or probability-based methods.
Rare event impairing long-term safety
Rare event impairing long-term safety shall refer to a potential event significantly reducing the performance of barriers that has a very low probability of occurring but that may jeopardize long-term safety in case of occurrence. Such an rare event may be caused by geological phenomena, climatic phenomena or human action.
- Rated value Rated value shall refer to the information used by the component manufacturer as a starting point for the design of a component or an I&C system platform, such as the nominal current of a breaker or the duration of a processing sequence of programmable logic. The manufacturer usually indicates the rated values in the component's brochure or specification.
- Real defect Real defect shall, in the context of qualification, refer to a defect that has developed in a structure during its manufacture or use without this having been advanced in any intentional fashion.
- Realistic defect Realistic defect shall, in the context of qualification, refer to a defect that has been intentionally caused in a test piece and that imitates the geometry of a real defect.
- Recording level Recording level shall refer to an indication threshold, above which indications shall be recorded in the inspection record.
- Redundancy Redundancy shall refer to the use of alternative (identical or diverse) structures, systems or system components, so that any one of them can perform the required function regardless of the state of operation or failure of any other.
- Reinforcement steel Reinforcement steel shall refer to steel used for non-prestressed reinforcement of a concrete structure.
Repair shall, in the context of Guide YVL A.8, refer to restoring the operability of a faulty SSC
- Repair work Repair work shall refer to the restoration of a faulty component or structure to a state which conforms to the original designs.
Representative person shall refer to an individual in the population group most highly exposed to a given radiation source (here to the radioactive substances contained in releases), whose radiation dose represents the doses received by such a population group (ICRP Publication 101). The representative person is equivalent to, and replaces, the previous term ‘average member of the critical group’.
- Research reactor Research reactor shall refer to a nuclear facility equipped with a nuclear reactor mainly used for the generation of neutron flux and ionising radiation for research and other purposes.
- Responsible manager Responsible manager shall refer to the person specified in Section 7 k of the Nuclear Energy Act.
- Review Review shall refer to activity undertaken to determine the suitability, adequacy and effectiveness of the measures needed to achieve set objectives.
Reviewing the effectiveness
Reviewing the effectiveness of operating experience feedback shall refer to assessing the functionality of the process. Reviewing the effectiveness of measures defined based on events shall refer to procedures to ensure that measures bring about the pursued plant-related or operational changes and that those measures that are insufficient for achieving the pursued changes are observed and responded to.
Risk analysis shall refer to examinations performed by using systematic measures in order to identify threats, problems, and vulnerabilities, surveying the causes and consequences thereof, and assess the related risks. (STUK Y/3/2016)
- Risk category Risk category shall refer to the magnitude of a risk caused by a pipe leak. The risk category is defined on the basis of the failure and consequence category, in turn defined based on the failure potential and the consequences of a failure.
- Risk management plan Risk management plan specifying the approach, the management components and resources to be applied to the management of risk. Risk management components typically include procedures, practices, assignment of responsibilities, sequence and timing of activities.
Risk matrix shall refer to a diagram consisting of three or more failure potentials (Y axis) and four or more consequence categories (X axis) describing the conditional core damage probability.
- Risk register Risk register shall refer to the records of identified risks, including results of analysis and planned responses.
- Risk-informed fire protection planning and assessment Risk-informed fire protection planning and assessment is realised when deterministic planning is combined with risk analysis while utilising demonstrative accident management methods.
- Risk-informed in-service inspection programme Risk-informed in-service inspection programme shall refer to an inspection programme where the selection process of non-destructive inspections as defined in Guide YVL E.5 is entirely based on the use of risk-informed methods.
- Risk-informed method Risk-informed methods shall refer to combining the results of a probabilistic risk assessment with the assessment of degradation mechanisms and the consequences of failure of components and structures.
- Risk-informed selection process Risk-informed selection process shall, within the context of non-destructive in-service inspections, refer to a selection process whereby the inspection areas, procedures and intervals of pressure-retaining piping segments are selected using risk-informed methods. A risk-informed selection process is applied in the preparation of the pre-service inspection plan, the in-service inspection programme for an inspection interval and the in-service inspection plan for operation period throughout the entire service life of a facility.
- Root causes Root causes shall refer to factors that explain the occurrence of direct causes. They indicate the suitable corrective action to be taken in order to avoid similar events. Root causes are those causes related to activities that can be identified and corrected with reasonable effort. In order for a cause to be considered a root cause, it must fulfil one of the following three conditions: the issue is replicable; the issue would not have occurred had this cause not existed; and the issue will not occur due to this cause if a correction is made.
Safe state shall refer to a state where the reactor has been shut down and is non-pressurised, and removal of its decay heat has been secured. (STUK Y/1/2019)
Safe state following a severe reactor accident
Safe state following a severe reactor accident shall refer to a state where the conditions for the controlled state of a severe reactor accident are met and, in addition, the pressure inside the containment is low enough that leak from the containment is minor, even if the containment is not leak-tight. (STUK Y/1/2018)
- Safety accessory Safety accessory shall refer to devices that are intended to protect pressure equipment against the allowable limits being exceeded. Safety accessories include: a) devices for direct pressure limitation, such as safety valves, bursting disc safety devices, buckling rods and controlled safety pressure relief systems; b) limiting devices which either activate the means for correction or provide for shutdown and lockout, such as pressure switches, temperature switches or fluid level switches and safety-related measurement, control and regulation devices.
Safety case shall refer to documentation for demonstrating compliance with the long-term safety requirements.
- Safety divisions Safety division shall refer to premises, physically separated from one another, and the components and structures contained therein, where one of the redundant parts of each safety system is placed.
Safety functions shall refer to functions important from the point of view of safety, the purpose of which is to control disturbances or prevent the generation or propagation of accidents or to mitigate the consequences of accidents (STUK Y/1/2018).
- Safety system Safety system shall refer to a system that has been designed to execute safety functions.
- Safety valve Safety valve shall refer to a valve type that automatically limits the pressure in a pressure vessel or system once the pressure or temperature exceeds the pre-set limits.
- Safety valve discharge coefficient Safety valve discharge coefficient shall refer to the ratio between the measured and theoretical flow capacity.
Safety-classified system/structure /component shall refer to a system, structure or component assigned to safety classes on the basis of its safety significance.
- Scenario Scenario shall refer to an evolution scheme describing the potential future behaviour of the disposal system.
- Secondary stress Secondary stress shall refer to stress generated in a pressure equipment structure as a result of the limiting of deformation, or to stress controlling the compatibility without incremental distortion (shakedown) of components whose temperatures and rigidity differ.
Security arrangements shall refer to the measures needed to protect the use of nuclear energy against illegal activities in the nuclear facility, its precincts other places or vehicles where nuclear energy is used. (Nuclear Energy Act 990/1987)
- Security guard Security guard shall refer to a person employed by a guarding services supplier and certified under Section 24 of the Private Security Services Act (282/2002), who has completed basic guard training or a corresponding qualification and meets the general requirements for the guard certification.
- Security organisation Security organisation shall refer to the work community consisting of the personnel designing, implementing or supervising nuclear security of a nuclear facility; the licensee's security organisation shall refer to a similar work community that is directly employed by the licensee. (Nuclear Energy Act 990/1987)
- Security plan Security plan shall refer to the account (preliminary security plan) presented in Section 35(6) of the Nuclear Energy Decree and the account presented in Section 36(7) of the Nuclear Energy Decree, as well as any changes to them.
- Security staff Security staff shall refer to persons trained and authorised to plan and implement nuclear security as laid down in Section 7 l of the Nuclear Energy Act.
- Security standing order Security standing order shall refer to a document referred to in Section 7 n of the Nuclear Energy Act.
- Security zone Security zone shall refer to the safety zone referred to in Section 4 of Government Decree 734/2008.
- Seismically classified system/structure/device Seismically classified system, structure or component shall refer to a system, structure or component which has been classified in different seismic categories based on the earthquake resistance requirements set for them.
Separation principle shall refer to physical separation and functional isolation. (STUK Y/1/2018)
Serially manufactured component
Serially manufactured component shall refer to a component which has not been designed particularly based on the customer’s specification but it is procured from an existing product line of the manufacturer. Typically one is manufactured in large quantities, and can be used for other applications, too. Functionality, structure, dimensions, materials, manufacturing process and quality of the component do not essentially differ within and between production lots.
Service life shall refer to the period of time beginning from the commissioning of the SSC fulfilling its operability requirements and ending when the degraded operability is not restored to the required level anymore.
- Service limit state dimensioning Service limit state dimensioning shall refer to dimensioning at a limit state where a structure no longer meets the requirements set as conditions for its usability (such as maximum crack width or displacement).
Severe accident shall refer to an accident in which a considerable part of the fuel in a reactor or the spent fuel in a fuel pool or storage loses its original structure. (Nuclear Energy Decree 161/1988)
Severe reactor accident
Severe reactor accident shall refer to an accident in which a considerable part of the fuel in a reactor loses its original structure. (STUK Y/1/2018)
Short-lived waste shall refer to nuclear waste, the calculated activity concentration of which after 500 years is below the level of 100 MBq/kg in each disposed waste package, and below an average value of 10 MBq/kg in waste in one emplacement room.
- Shut down reactor Shut down reactor shall refer to a reactor in a subcritical state with an effective multiplication factor, taking uncertainties into consideration, of less than 0.995.
Significant function with respect to safety
Significant functions with respect to safety shall refer to functions in accordance with the Radiation and Nuclear Safety Authority Regulations STUK Y/1/2018 and STUK Y/4/2018. The licensee shall designate these functions and verify the competence of the persons performing these functions. Functions in accordance with Sections 7 i and 7 k of the Nuclear Energy Act (990/1987) are always considered significant functions with respect to safety. Approval of the individuals proposed to these functions shall be requested from STUK.
- Single failure Single failure shall refer to a failure due to which a system, component or structure fails to deliver the required performance.
- Single failure criterion Single failure criterion (N+1) shall mean that it must be possible to perform a safety function even if any single component designed for the function fails.
Site area shall refer to an area in use by nuclear power plant units and other nuclear facilities in the same area, and to the surrounding area, where movement and stay are restricted by the Decree of the Ministry of the Interior issued under Section 52 of the Police Act (493/1995). (STUK Y/2/2018)
- Site pursuant to the Additional Protocol to the Safeguards Agreement Site pursuant to the Additional Protocol to the Safeguards Agreement shall refer to an area delimited in the relevant design information (Basic Technical Characteristics) for a facility (including closed-down facilities), the limits of which have been defined by the Radiation and Nuclear Safety Authority in collaboration with the European Commission following consultations with the operator.
Small-diameter piping shall refer to piping with DN ≤ 50.
- Software to be developed Software to be developed during the present design work, such as software for a programmable logic application.
- Software-based Software-based shall refer to a function achieved by means of executing program code with a processor. The term should not be confused with the concept of programmable. For example, a software-based temperature transducer is in no way “programmable” after manufacture, but its operation is based on program code executed by a processor.
Source material shall refer uranium that contains the mixture of isotopes in amounts occurring in nature, or uranium in which the relative quantity of isotope 235 has been reduced, thorium, and any of the materials specified above in the form of metal, alloy, chemical compound or concentrate. (161/1988, Section 3)
Spare part shall refer to a back-up part for an SSC that can be used to restore the reduced or lost operability to the required level.
- Special arrangements Special arrangements shall refer to STUK-approved measures enabling the transport of consignments that do not fully meet the requirements for radioactive materials set forth in the transport regulations for dangerous goods.
- Special fissionable material Special fissionable material shall refer to plutonium-239, uranium-233, uranium enriched in isotopes 235 and 233, and materials that which contain one or more of the above-mentioned materials (161/1988, Section 3).
Special processes shall refer to manufacturing processes, the results of which cannot be directly verified by means of a product inspection or testing after manufacture; instead, any shortcomings in the process may only appear later while the product is in use. Special processes include, for instance welding, forming and heat treatment.
- Special situation Special situation shall refer toan exceptional situation such as an accident, deterioration of nuclear security arrangements , a disturbance or a threat situation.
- Specimen Test specimen shall refer to a piece manufactured for testing from a concrete, steel or reinforcement sample.
- Spent nuclear fuel Spent nuclear fuel shall refer to nuclear material that has been used as nuclear fuel for the production of nuclear energy and that contains significant amounts of nuclear waste. (Nuclear Energy Decree 161/1988)
- Standardised component Standardised component shall refer to a component dimensioned in accordance with the standards applied that is chosen for the intended location of use based on the determined value of its nominal size, pressure rating class, property class etc.
- Steel structure Steel structure shall refer to structural steel componentss that are delivered as construction products. Typical steel structures of nuclear facilities include: load-bearing structures of buildings, load-bearing steel structures of the concrete reactor containment, vessels subject to hydrostatic pressure, piping break supports, missile protectors, storage racks for fresh and spent fuel, gates and linings of spent fuel pools, crane rail supports, doors and hatches, steel platforms and spent fuel handling equipment and crane rails.
- Strategic spare part (hoisting equipment) Strategic spare parts shall refer to spare parts the availability of which ensures the operability of the hoisting device unit.
- Strength analysis Strength analysis shall refer to determination of stresses and deformations in the structure using the load provided (or other stress), or to determination of the maximum loads using the allowable stresses and deformations when the geometry, dimensions and material of the structure are known; strength analyses also include the loading analyses stress analyses, fatigue analyses, brittle fracture analyses and leak before break (LBB) analyses.
- Strength analysis report Strength analysis report shall refer to a set of documents consisting of the pressure equipment strength analysis documentation submitted for regulatory review by authorities in connection with the construction plan.
- Stress analysis Stress analysis shall refer to a strength analysis based on the modelling of the actual structure and loads of pressure equipment,which is used to eliminate the risk of failure caused by the loss of the load-bearing capacity, excessive deformation and fatigue, when the acceptance limits that have been set for the calculated stresses governing these mechanisms, as stipulated in the applicable standard, are met.
- Stress analysis (hoisting equipment) Stress analysis shall, in the context of Guide YVL E.11, refer to the calculation of stresses and displacement applying the rules of the strength theory so that external forces, displacement as well as pressure stresses and temperature stresses do not result in exceeding the allowable values.
- Structural analyses Structural analyses shall refer to dynamic analyses and stress analyses.
- Subcontractor Subcontractor shall refer to a supplier that is not in a direct contractual relationship with the licensee or licence applicant.
- Subcritical state Subcritical state shall refer to a state where no chain reaction sustained by neutrons released by nuclear fission occurs.
- Suitability assessment A suitability assessment presents how well a fire protection system meets the requirements placed on it and how the licensee has verified conformity. The suitability assessment also lists changes to the approved documents and their effect on the suitability and acceptability of the system in question.
- Summary of justifications Summary of justifications shall refer to a document that presents how the structure meets the requirements set for it, and how the licensee has established its conformity to requirements. The summary of justifications shall also present the changes made to the approved documents, any non-conformances that occurred during manufacture, and their impact on the suitability and acceptability of the structure.
Summary programme for in-service inspections
Summary programme for in-service inspections shall refer to a summary programme required under Section 36 of the Nuclear Energy Decree (161/1988).
- Supervised area Supervised area shall refer to an area where working conditions are supervised in order to protect employees against radiation. However, area is not designed as controlled area and normally no special radiation protection measures are needed.
Supplier shall refer to an organisation or person manufacturing or providing a product or service. (SFS-EN ISO 9000:2015)
- Supplies Supplies shall refer to the chemicals, sealing materials, insulation, plastics, grinding discs and other similar materials and items that the various power plant organisations use in their activities. Not all of the above products are among safety-classified supplies at a power plant, if they can be categorised as component-specific spare parts with system- and component level requirements.
- Supply chain Supply chain shall refer to all resources, processes and functions that are required to create products. These may include the design of the product, and the manufacture, assembly and delivery of material and components.
- System System shall refer to a combination of components and structures that performs a specific function.
- System (information security) A system in the context of information security shall refer to a system consisting of persons, information processing equipment, data transfer equipment and software intended to intensify or facilitate a certain function or to make it possible. The system may be an infor¬mation system, a communications system, an electrical or I&C system, or a communication system for security surveillance or emergency preparedness.
- System performance test System performance tests shall refer to tests that ensure that systems and their components function as planned. System performance tests include both the testing of single systems and the joint functional tests of multiple systems.
- System platform software System platform software shall refer to component software that is not modified for different locations of use, such as the real-time operating system, network communication routines, or function libraries. Generally, users cannot access the system platform software.
System/structure/device important for safety
System/structure/component important to safety shall refer to systems, structures or components in safety classes 1, 2 and 3 and systems in class EYT/STUK.
- Systematic failure Systematic failure shall refer to failure that is not random failure.
Systems, structures and components (SSC)
Systems, Structures and Components (SSC) shall refer to mechanical, electrical, I&C or civil system, structure or component in safety classes 1, 2 and 3 or in class EYT/STUK (non-nuclear).
- Technical justification Technical justification as part of qualification for non-destructive in-service inspections shall refer to evidence establishing that an inspection system can meet the objectives set for it.
- Temporary modification Temporary modification shall refer to a provisional solution replacing an accepted solution pertaining to systems, structures and components or related methods and procedures that cannot be properly corrected without delay using established procedures.
- Testing Testing shall refer to determining one or more characteristics of an object evaluated for conformity to requirements. (SFS-EN ISO/IEC 17000, 2005)
- Testing body approved by STUK Testing body approved by STUK shall refer to a testing organisation that has been approved in accordance with Guide YVL E.1.
Testing organisation shall refer to an organisation performing testing activities requiring special competence. (Nuclear Energy Act 990/1987)
Third party shall refer to an individual or organisation that is independent of the individuals or organisations responsible for the design, manufacturing, supply, installation, procurement, ownership, operation or servicing of the item analysed.
Threat shall refer to a situation in which unlawful action against a nuclear facility, nuclear material, or nuclear waste is ascertained, or there is reason to suspect this. (STUK Y/3/2016)
Time limited ageing qualification or analysis
Time limited ageing qualification or analysis shall refer to testing or calculational analysis to demonstrate the service life of SSC when retaining of the operability is limited in time as the integrity or functionality of the SSC will degrade with relation to duration of stressors acting on the SSC in operating conditions.
- Transfer cask Transfer cask shall refer to a container in which spent nuclear fuel is transported within the plant site.
- Transition temperature Transition temperature shall refer to a temperature, which characterises an intrinsic substantial ductile-to-brittle transition of ferritic steels with decreasing temperature, and is determined through destructive testing in compliance with the applicable standard.
- Transport Transport shall refer to an event specified in Section 3(1)(2) and Sections 3(1)(4)–(7) of the Act on the Transport of Dangerous Goods (719/1994) concerning nuclear material or nuclear waste.
- Transport index Transport index shall refer to a figure marked on the warning labels and transport documents of a package indicating the dose rate of radiation emitted from the package.
- Trend analysis Trend analysis shall refer to the procedures to identify degradation or undesired developments by means of analysing events or failures that have occurred.
Trickle charge voltage for a set of accumulators
The trickle charge voltage for a set of accumulators shall refer to a voltage generated by a battery charger in order to maintain a charge level on a set of accumulators that exceeds the nominal voltage of the set of accumulators.
- Type approval Type approval is granted by a type approval organisation or the Ministry of the Environment for a particular reason (Act on the type approval of certain construction products [954/2012]).
- Type approval (building products) Type approval for building products shall refer to a decision demonstrating that the characteristics of a building product meet the essential technical requirements set forth in the Land Use and Building Act (132/1999) or those decreed by virtue of the Act.
- Type approved pump Type approved pump shall refer to a pump, the conformity to requirements of which a third party verifies by applying the modules of Decision 768/2008/EC.
- Type approved valve Type approved valve shall refer to a valve, the conformity to requirements of which a third party verifies by applying the modules of Decision 768/2008/EC.
- Type test Type test shall refer to a test whose results can be used to ensure that the machine unit, or a part thereof, has been designed to meet the usage location-specific operability requirements.
- Ultimate limit state dimensioning Ultimate limit state design shall refer to design at a limit state where the structure is considered to lose its load-bearing capacity in part or entirely.
Unlawful action shall refer to an activity or measure that is aimed at directly or indirectly endangering the nuclear safety or radiation safety of a nuclear facility, nuclear material or nuclear waste. Deliberate or negligent activity that is punishable by law towards a nuclear facility, nuclear material or nuclear waste, or towards the persons working at a nuclear facility is considered unlawful action. (STUK Y/3/2016)
- Unprocessed waste Unconditioned waste shall refer to waste that has not been processed into its final form or packed into a waste package for storage and/or disposal.
Use of nuclear energy
Use of nuclear energy shall refer to the operations laid down in Sections 2(1) and 2(2) of the Nuclear Energy Act. (Nuclear Energy Act 990/1987)
According to Section 2(1), the Act applies to:
1) the construction, operation and decommissioning of nuclear facilities;
2) mining and enrichment operations aimed at producing uranium or thorium;
3) the possession, manufacture, production, transfer, handling, use, storage, transport and import of nuclear material;
4) the possession, manufacture, production, transfer, handling, use, storage, transport, export and import of nuclear waste;
4 a) the disposal of nuclear waste in small-scale as opposed to large-scale
5) in cases to be provided for by Government decree, the possession, manufacture, assembly, transfer and import of the following material, devices, equipment, or information, should they prove pertinent to the proliferation of nuclear weapons or should the obligations under Finland’s international treaties in the field of nuclear energy have a bearing on them:
a) non-nuclear material, when its properties are particularly suited to creating nuclear energy;
b) devices and equipment intended or otherwise particularly suited for use in nuclear facilities;
c) devices and equipment intended or otherwise particularly suited for use in the manufacture of nuclear material or material referred to in item a);
d) equipment essential for the manufacture of the devices or equipment referred to in items a) and b)
e) nuclear information that is in written or some other physical form and not generally available
6) export and import of ores containing uranium or thorium, to be specified under Government decree.
According to Section 2(2) of the Nuclear Energy Act, the application of the Nuclear Energy Act shall be provided for by Government Decree, with respect to:
1) the conclusion and implementation of a private law agreement, for implementation outside Finland in regard of any of the activities referred to in this section (Section 2), with a foreign State, a foreign person or corporation, should the agreement prove pertinent to the proliferation of nuclear weapons or should the obligations under international treaties in the field of nuclear energy, to which Finland is a Party, have a bearing on the agreement; and
2) Nuclear fuel cycle-related research and development activities determined in Article 18(a) of the Protocol Additional (53/2004) to the agreement made on the implementation of Article III (1) and (4) of the Treaty on the Non-Proliferation of Nuclear Weapons between Austria, Belgium, Denmark, Finland, Germany, Greece, Ireland, Italy, Luxembourg, the Netherlands, Portugal, Spain, Sweden, the European Atomic Energy Community and the International Atomic Energy Agency.
(Nuclear Energy Act 990/1987)
User inspectorate shall refer to an inspection organisation as referred to in Article 16 of Pressure Equipment Directive 2014/68/EU.
- Utilisation rate Utilisation rate shall refer to the ratio between the design load of the structure and the capacity of the structure
Validation shall refer to confirmation, through the provision of objective evidence, that the requirements for a specific intended use or application have been fulfilled.
- Valve Valve shall, in the context of Guide YVL E.8, refer to an entity (valve unit) consisting of the valve, its actuator and pilots, and any other structures and parts that are essentially involved in the entity. Whenever a specific part of a valve is discussed in Guide YVL E.8, it is mentioned separately.
- Valve flow factor Valve flow factor shall refer to a measured constant that can be used to calculate the volumetric flow of the valve when the pressure difference across the valve is known.
Ventilation shall refer to maintaining and improving the quality of indoor air by circulating it; in some rooms of a nuclear facility, ventilation systems are also used to limit the spread of radioactive substances.
- Ventilation systems Air conditioning systems shall refer to systems designed to manage the purity, temperature, humidity and movement of indoor air by treating supply air or circulating air.
- Verification Verification shall refer to confirmation, through the provision of objective evidence, that set requirements have been fulfilled.
- Verification certificate Verification certificate shall refer to a certificate verifying that the characteristics of a building product are in line with the Land Use and Building Act (132/1999) or the provisions issued by virtue of the Act for the intended purpose of use defined by the manufacturer. The verification certificate is granted by a body approved by the Ministry of the Environment (Act on the Approval of Certain Construction Products [954/2012]).
Very low-level waste
Very low-level waste shall refer to nuclear waste whose average activity concentration of significant radionuclides does not exceed the value of 100 kBq per kilogram and the total activity does not exceed the values laid down in Section 6(1) of the Nuclear Energy Decree (161/1988).
- Vital area Vital area shall refer to to locations and premises within the protected area of a nuclear facility from which it is possible to conduct unlawful actions that may lead to significant radiation consequences, for which reason these locations and premises must be placed under special protection.
- Waste package Waste package shall refer to the entity formed by waste and the container surrounding it.
- Welding filler material Welding filler materials shall refer to the filler wires, covered electrodes and other similar welding consumables used in welding.
- Welding material Welding consumables shall refer to welding filler materials and auxiliary welding materials used in welding.
- Witness point Witness point shall refer to an inspection for which advance invitations have been sent to the parties defined in the inspection plan but whose supervision is not a condition for proceeding with the work. Having received the invitation, the invited parties may, however, separately require that they be present in order for the work to be continued.
- Works test Works test shall refer to a test carried out by persons performing the work, in conditions equivalent to production conditions and prior to the commencement of the work.
- Worst case defect Worst case defects shall, in the context of qualification, refer to defects, component geometry or other essential parameters that are likely to pose the greatest challenges for defect detection and the accurate determination of size within the framework of a specific situation, as defined by the input information, using a set inspection system.
- Zone classification Zone classification shall refer to the division of the premises of the controlled area into zones based on the external dose rate, surface contamination and airborne radionuclide concentration.