SSG-90 Radiation Protection Aspects of Design for Nuclear Power Plants

IAEA Safety Standards Series No. SSR‑2/2 (Rev. 1), Safety of Nuclear Power Plants: Commissioning and Operation [5];

IAEA Safety Standards Series No. SSR‑1, Site Evaluation for Nuclear Installations [6];

IAEA Safety Standards Series No. GSR Part 2, Leadership and Management for Safety [7];

IAEA Safety Standards Series No. GSR Part 4 (Rev. 1), Safety Assessment for Facilities and Activities [8];

IAEA Safety Standards Series No. GSR Part 5, Predisposal Management of Radioactive Waste [9];

IAEA Safety Standards Series No. GSR Part 6, Decommissioning of Facilities [10];

IAEA Safety Standards Series No. GSR Part 7, Preparedness and Response for a Nuclear or Radiological Emergency [11].

Describes the safety requirements that are applicable to radiation protection, including those relating to the principles of dose limitation and optimization, as a basis for the measures to be implemented in the design of nuclear power plants;

Provides recommendations for the measures to be taken in the design for the protection of site personnel, the public and the environment;

Outlines the methodologies used to calculate on‑site and off‑site radiological conditions and to verify that the design provides an adequate level of radiation protection.

To prevent accidents with harmful consequences resulting from a loss of control over the reactor core or over other sources of radiation, and to mitigate the consequences of any accidents that do occur;

To ensure that for all accidents taken into account in the design of the installation, any radiological consequences would be below the relevant limits and would be kept as low as reasonably achievable;

To ensure that the likelihood of occurrence of an accident with serious radiological consequences is extremely low and that the radiological consequences of such an accident would be mitigated to the fullest extent practicable.”

The choice of materials, so that amounts of radioactive waste will be minimized to the extent practicable and decontamination will be facilitated;

The access capabilities and the means of handling that might be necessary;

The facilities necessary for the management (i.e. segregation, characterization, classification, pre‑treatment, treatment and conditioning) and storage of radioactive waste generated in operation, and provision for managing the radioactive waste that will be generated in the decommissioning of the plant.”

The effects of natural and human induced external events occurring in the region that might affect the site;

The characteristics of the site and its environment that could influence the transfer of radioactive material released from the nuclear installation to people and to the environment;

The population density, population distribution and other characteristics of the external zone, in so far as these could affect the feasibility of planning effective emergency response action [11], and the need to evaluate the risk to individuals and to the population.”

Suitable assembly points, provided with continuous radiation monitoring;

A sufficient number of suitable escape routes;

Suitable and reliable alarm systems and other means for warning and instructing all persons present under the full range of emergency conditions.”

Radiation exposure should be taken into account early in the design process and should be reduced by means of radiation protection measures such that further expenditure on design, construction, operation and decommissioning would not be warranted by the associated reduction in radiation exposure.

Measures such as reducing major disparities in the occupational doses received by workers of different types who work within the controlled area and avoiding arduous working conditions in radiation areas should be taken into account in the design.

GSG‑7 [3], which addresses the radiation protection of workers;

IAEA Safety Standards Series No. GSG‑8, Radiation Protection of the Public and the Environment [25];

IAEA Safety Standards Series No. GSG‑3, Safety Case and Safety Assessment for the Predisposal Management of Radioactive Waste [26], which addresses wider aspects of optimization of protection and safety for predisposal management facilities;

IAEA Safety Standards Series No.GSG‑9, Regulatory Control of Radioactive Discharges to the Environment [27], which provides recommendations on the application of the principles of radiation protection and the safety objectives associated with the control of discharges and on the process for the authorization of discharges;

GSG‑10 [20], which describes a framework and methodologies for prospective radiological environmental impact assessment, covering the assessment of public exposure only.

The reactor core (including fuel and non‑fuel components), reactor internals and vessel, and the surrounding materials that are activated;

The reactor coolant and associated systems such as the moderator, volume control and reactor water cleanup systems;

The steam supply system, feedwater system and turbine generators (depending on the design);

The waste treatment systems, spent resins and storage systems;

Irradiated fuel, including connected cooling and cleaning systems;

The fuel handling and storage system;

Decontamination facilities;

Ventilation systems;

Miscellaneous sources such as sealed sources that are used for non‑destructive testing.

Knowledge of the practices that result in the exposure of site personnel and members of the public;

Feedback of operating experience to the design team;

Familiarity with the main factors that influence individual doses and collective dose;

Familiarity with the analytical methods that are available to assist in the optimization of the design;

Recognition that specialists in radiation protection are to be consulted whenever necessary to ensure that aspects that will have implications for radiation protection are properly evaluated and taken into account in the design.

Expertise in all areas that affect the production of radioactive material and its transport and accumulation in the plant and the dispersion of radionuclides in the environment;

The evaluation of the different sources of radiation in the plant and the resulting doses using the best available analytical methods and data from relevant operating experience;

Familiarity with the relevant regulations, guidance and best practices;

Familiarity with maintenance, in‑service inspection and other work in areas in which there are high radiation levels and that make a major contribution to the radiation exposure of site personnel.

Rules for the layout of the plant;

Design measures to minimize the use of personal protective equipment;

Checklists⁶ that can be used by engineers and reviewed by radiation protection officers.

Radiation protection officers within the design organization should be consulted at the early stages of the design when the options for the major aspects of the design are being evaluated. It may also be appropriate to consult with specialists from external organizations.

The design should incorporate good engineering practices that operating experience has shown to be effective in reducing exposure; deviations from such practices should be accepted only when a net benefit has been demonstrated.

Radiation protection officers should review all decisions that might have a major influence on exposures.

There should be an appropriate ongoing forum for proposing improvements and resolving disputes that might occur between design engineers and radiation protection specialists.

Annual collective dose targets and individual dose targets (e.g. for average and maximum effective doses) for site personnel;

Annual individual dose targets for members of the public.

The general criteria for the plant design should be developed and documented. These criteria should include the principles on which the layout of the plant will be based and restrictions on the use of specific materials in the design of the plant. The documents produced will form part of the management system for the design (see IAEA Safety Standards Series No. GS‑G‑3.5, The Management System for Nuclear Installations [40]).

A strategy for controlling exposures should be developed so that the most important aspects are considered early in the design and in a logical order.⁷ Consequently, these aspects should be considered first in the design stage, and it should be ensured that the design has been proven. The aim should be to reduce exposures to levels that are as low as reasonably achievable, which will also help to improve the availability and therefore the economic performance of the plant. Design features that minimize the production and buildup of radionuclides should also be considered; such features will reduce radiation and contamination levels throughout the plant, whereas a local solution (e.g. increasing the shielding, improving ventilation) will have only a local benefit. Local plant features such as the plant layout, the shielding and the design of systems and components, should subsequently be considered. An example of a simplified strategy for reducing exposures in a pressurized water reactor is shown in Fig. 2.

A logical layout for the plant should be developed, dividing the plant into zones on the basis of predicted dose rates and contamination levels (see para. 6.73 of SSR‑2 (Rev. 1)), access requirements and specific issues such as the need to separate safety trains⁸. The dose rates may be calculated using the source terms that form the basis for the radiation protection aspects of design (see Annex I), or they may be based on operating experience from similar plants, provided that any differences in the design and operating parameters are not significant. The zoning should be consistent with national legislation and regulatory requirements. It may be adequate to use the same zones that will be used when the plant is operating, but a more specific definition of the zones is often necessary for design purposes (see the examples given in Annex II).

The maintenance programme and operational tasks should be defined, preferably on the basis of well established concepts. The number of staff for each task should be based only on the operational and security requirements and should not be artificially increased, to comply with the regulatory requirements or the dose constraints. For tasks for which doses are predicted to be relatively minor, the work can be expressed generically in terms of the number of person‑hours that will be spent in each radiation zone. The type of worker who will perform each task should also be identified, for example maintenance personnel, in‑service inspection personnel, technical support personnel (e.g. scaffolders, insulation workers), decontamination staff and radiation protection officers.

Collective doses and individual doses should be evaluated by combining the results of steps 3 and 4. The use of a database on occupational exposure is recommended. Maximum use should be made of relevant operating experience, where available, particularly for work that is difficult to predict, such as unplanned maintenance.

The proposed procedure is shown in Fig. 3, which presents a schematic flow chart of the factors that determine individual and collective doses. This procedure should be repeated at each significant stage of the design, and the level of detail should increase as the design is developed. At each stage, the doses that are evaluated should be compared with the design targets for each type of work. The factors that determine individual and collective doses are shown in Fig. 3. At each step in Fig. 3 involving design options, optimization studies should be performed. This is particularly important in cases for which it is predicted that the design targets will be exceeded.

The hierarchy of control measures should be considered carefully. Hazard elimination is the best solution, followed by hazard reduction if elimination is not possible. The next step should be to isolate the hazard and to use engineered controls to reduce the risk to workers, with administrative controls and personal protective equipment at the bottom of the hierarchy.

Reduction of dose rates in working areas by the following:

Reduction of occupancy times in radiation fields by the following:

Reducing sources (e.g. by the appropriate selection of materials; reduction of surface and airborne contamination; decontamination measures; control of corrosion, water chemistry, filtration and purification; and the exclusion of foreign material from the primary systems);

Improving shielding;

Increasing the distance between workers and sources (e.g. through remote handling);

Controlling the direction of air flow containing radionuclides and improving filtered ventilation, especially in pressurized heavy water reactors.

Specifying high standards of equipment to ensure very low failure rates;

Ensuring ease of maintenance and of equipment removal;

Removing the necessity for certain operational tasks (e.g. by providing built‑in auxiliary equipment or making provision in the design for permanent access);

Ensuring ease of access and good lighting;

Optimizing the number of workers and their time in a radiation field by design means;

Making provisions for remote or automated inspection devices;

Isolating sources and important passages.

Site specific features that affect doses to members of the public should be identified at an early stage of the design process and taken into account in the design (see GSG‑10 [20]). This should include the identification of the representative person and the exposure pathways, which should be subject to the approval of the regulatory body. One possible approach would be to set targets for radioactive releases in which account is taken of operating experience and the use of best practicable means in the design of the treatment systems for radioactive effluents.

The resulting doses to the representative person should be evaluated to ensure achievement of the target.

The risks to the public from possible releases of radioactive material from the nuclear power plant;

The risks to site personnel from these releases and from direct radiation exposure.

Sources for which the design of the shielding will be determined;

Sources for which shielding is not practicable and that may be major sources of doses to workers during plant operation;

Sources that are major sources of doses to workers during decommissioning;

Sources that present special hazards to workers during plant operation, such as particles containing alpha emitters or high concentrations of activated cobalt;

Sources that are important contributors to doses to members of the public during plant operation;

Sources that are important contributors to doses to members of the public during plant decommissioning.

Corrosion products;

Fission products;

Activation products.

Particles of metal resulting from excessive wear of components or fuel assemblies;

Debris left in the primary circuit or other circuits connected to it;

Pieces of crud deposits on the fuel.

Carbon‑14, ³H and ⁸⁵Kr, because the best practicable means available for their removal by waste treatment systems are not efficient and because they have long half‑lives;

Argon‑41, which is an important contributor even though it has a short half‑life, because it is released in large volumes of air (e.g. in venting of the containment during operation for some pressurized water reactors);

Xenon‑133, which is a weak gamma emitter, but it may be of importance when the reactor has been operating with a significant number of defects in the fuel cladding;

Iodine, caesium and corrosion products.

Provision of passageways of adequate dimensions for ease of access to plant systems and components. In areas where it is likely that site personnel will have to wear full protective clothing, including masks with portable air supplies or connections to a supply by air hoses, account should be taken of this in deciding on the dimensions of the passageways.

Provision of clear passageways of adequate dimensions to facilitate the removal of plant items to a workshop for decontamination and repair or disposal.

Provision of adequate space in the working areas to perform tasks such as repairs or inspections.

Provision of access to high radiation areas such as pressurized water reactor steam generators, and valves in systems that contain primary coolant.

Provision of ‘waiting areas’ in low radiation areas.

Placement of components that are likely to be operated frequently or need maintenance or removal at a convenient height for working.

Remote or automated operation for components that will be operated frequently.

Provision of ladders, access platforms, crane rails or cranes in areas where their use is foreseen for the maintenance or removal of plant components. Features to facilitate the installation of temporary shielding should be included in the design.

Use of computer aided design models to optimize design aspects that affect working times. Video or photographic records should be made during the construction of the plant to facilitate the planning of work in areas of high radiation levels during operation and thus to shorten working times.

Provision of means for the quick and easy removal of shielding and insulation where this is necessary in order to perform routine maintenance or inspection.

Provision of special tools and equipment to facilitate work and thus reduce exposure times.

Provision of remotely controlled equipment.

Provision of a suitable system for communication with the site personnel working in radiation areas or contamination areas.

Provision of electrical supplies that are easy to disconnect from and reconnect to equipment, and sufficient availability of sockets.

Control of local areas during transfer of high activity radiation sources so as not to affect other simultaneous operations.

The workspace around components that need regular maintenance in areas of high radiation levels should be shielded from the radiation from other systems. Where possible, components that need regular maintenance should not be placed in such areas.

Non‑radioactive components that do not have to be mounted close to active components should be installed outside areas of high radiation levels.

Methods for sampling radioactive liquids that involve minimal exposure should be provided. Automated methods should be used where reasonably achievable.

Methods for dose reduction (e.g. flushing) to avoid the sedimentation of radioactive sludge in piping and containers should be provided.

For mutual standby systems, if it is necessary to maintain one system while the other system is in operation, sufficient shielding should be installed between the two systems.

Minimizing the area and number of straight‑through paths containing material of very low density (e.g. gases, including air).

Placing the penetrations in areas of low occupancy or no personnel access in the shine path (e.g. at height).

Providing shielding plugs.

Providing zigzag or curved pathways to guarantee no line of sight through shields. In this case, the width or density of the shielding may be increased near the penetration to compensate for the loss of material due to the penetration.

Filling the gaps with grouting or other compensatory shielding material.

Mechanisms of thermal and mechanical mixing;

The limited effectiveness of dilution in reducing airborne contamination;

Exhausting of the air from areas of potential contamination at points near the source of the contamination;

The use of exhaust rates that are commensurate with the potential for contamination in the area;

The need to ensure that the exhaust air discharge point is not close to an intake point of the ventilation system;

The need to avoid extracting potentially contaminated air in the vicinity of likely worker positions.

Maintaining the confinement of stored materials;

Providing for radiation protection (by means of shielding and contamination control);

Providing for ventilation, as necessary;

Providing for inspection and/or monitoring of the waste packages and storage facility, as necessary;

Allowing for maintenance and repair of waste packages;

Allowing for retrieval of radioactive waste for transport off the site;

Allowing for expansion of the storage capacity, as appropriate;

Allowing for movement of waste inside the storage facility;

Considering eventual decommissioning.

To prevent criticality by a specified margin, by physical means or by means of physical processes, and preferably by use of geometrically safe configurations, even under conditions of optimum moderation;

To permit inspection of the fuel;

To permit maintenance, periodic inspection and testing of components important to safety;

To prevent damage to the fuel;

To prevent the dropping of fuel in transit;

To provide for the identification of individual fuel assemblies;

To provide proper means for meeting the relevant requirements for radiation protection;

To ensure that adequate operating procedures and a system of accounting for, and control of, nuclear fuel can be implemented to prevent any loss of, or loss of control over, nuclear fuel.”

To permit adequate removal of heat from the fuel in operational states and in accident conditions;

To prevent the dropping of spent fuel in transit;

To avoid causing unacceptable handling stresses on fuel elements or fuel assemblies;

To prevent the potentially damaging dropping of heavy objects such as spent fuel casks, cranes or other objects onto the fuel;

To permit safe keeping of suspect or damaged fuel elements or fuel assemblies;

To control levels of soluble absorber if this is used for criticality safety;

To facilitate maintenance and future decommissioning of fuel handling and storage facilities;

To facilitate decontamination of fuel handling and storage areas and equipment when necessary;

To accommodate, with adequate margins, all the fuel removed from the reactor in accordance with the strategy for core management that is foreseen and the amount of fuel in the full reactor core;

To facilitate the removal of fuel from storage and its preparation for off‑site transport.”

Chemical stability against corrosion caused by processes acting within the waste and/or caused by external conditions;

Protection against radiation damage, especially stability under conditions of the degradation of organic materials and damage to electronic devices;

Resistance to impacts caused by operational loads, incidents or accidents;

Resistance to thermal effects, if applicable;

If the waste contains nuclides emitting high‑energy beta rays (e.g. ⁹⁰Y), bremsstrahlung in vessels made of materials containing elements with a high atomic number.

Generation of hazardous gases caused by chemical and radiolytic effects (e.g. the generation of hydrogen gas caused by radiolysis) and the buildup of overpressure;

Generation of combustible or corrosive substances;

Acceleration of the corrosion of metals (in particular, mild steel).

A health physics operations office, including storage and testing facilities for radiological instruments and protective equipment, facilities (with sufficient space, and power and gas supplies) to take sample measurements, and computer equipment for the display of remote monitoring data;

A changing room for changing into and out of protective clothing;

A personnel decontamination facility including provisions for showering and hair washing;

An equipment decontamination facility;

Laundry facilities for contaminated clothing, where these services are not provided by an external provider;

A first aid room;

A radiochemistry laboratory for the preparation of samples and the measurement of activity, the specifications of which are in accordance with the reactor chemistry analysis needs;

A storage area for contaminated items and tools;

A workshop for maintenance of contaminated equipment;

A secure storage location for radiation sources;

Facilities for the management and storage of waste;

An effluent storage tank area;

A dosimetry laboratory, or a dosimetry control point if this service is provided by an external provider;

A data recording and storage system for creating relevant databases (e.g. for dosimetry records or instrument control) and for updating them as necessary;

An alternative or remote health physics control centre;

Assembly areas at the plant for use during an emergency;

Emergency response facilities;

An identified sheltering area for plant personnel.

Protective clothing and boots;

Respiratory protective equipment;

Air samplers and other equipment for measuring airborne activity concentrations;

Portable dose rate meters with an audible alarm at variable settings, and devices for monitoring personnel contamination and surface contamination;

Portable shielding, signs, ropes, stands and remote handling tools;

Communication equipment;

Meteorology instruments;

Equipment for monitoring individuals for intakes of radionuclides;

Temporary containers for solid radioactive waste and special containers for radioactive liquids;

Emergency equipment (including extendable high dose rate monitors, additional protective clothing, self‑powered air samplers and emergency vehicles);

First aid equipment;

Equipment for sampling and analysis around waste storage areas, such as borehole monitoring equipment for underground storage facilities for radioactive waste;

Personal dosimeters for monitoring individual external exposure.

The effects of events that might happen in nuclear installations located in the area surrounding the site (see also SSG‑68 [15]);

Site characteristics that are relevant in the transfer of radioactive material potentially released from the nuclear power plant, including site characteristics relevant to the design of the escape routes aiming to minimize the risk to site personnel and the public during evacuation (see NS‑G‑3.2 [19]);

Density, distribution and habits of the population (see NS‑G‑3.2 [19]) that are relevant to evaluation of the risk for individuals and for the population as a whole, and other characteristics of the area surrounding the nuclear installation that could affect the feasibility of planning effective emergency response actions (see SSR‑1 [6]);

At multiple unit sites, the effects of events in one nuclear power plant unit on one or more other units (see IAEA Safety Standards Series No. SSG‑79, Hazards Associated with Human Induced External Events in Site Evaluation for Nuclear Power Plants [64]).

Suitable assembly points, provided with continuous radiation monitoring;

A sufficient number of suitable escape routes;

Suitable and reliable alarm systems and other means for warning and instructing all persons present under the full range of emergency conditions.”

The assessment, control and recording of the doses received by emergency workers;

Procedures to ensure contamination control;

The provision of appropriate personal protective equipment (which depends on the severity of the hazard), procedures and training for emergency response in the postulated hazardous conditions;

Arrangements for the provision and secure storage of sufficient amounts of consumables (e.g. respiratory protection and protective clothing) that will be necessary during the postulated event.

Ensuring that event sequences and accident scenarios potentially leading to an early radioactive release or a large radioactive release are ‘practically eliminated’ (see para. 5.27 of SSR‑2/1 (Rev. 1) [1] and the recommendations provided in SSG‑88 [21]);

Providing design means to minimize the scope of fuel damage and protect the barriers against releases of fission products from the fuel;

Providing design means to maximize the integrity of primary system pipework;

Achieving leaktightness, isolation and bypass prevention of the containment;

Ensuring adequate safety features are provided to condense steam and/or maintain containment pressures within design limits following all accidents including severe accidents;

Filtering the exhaust air or using charcoal delay beds in order to reduce the releases of airborne radioactive material, with due account taken of the fact that some pathways for accidental releases may bypass the filtered exhaust system;¹⁸

Achieving a high decontamination factor for the filters by using best practices in the design, the filter material and the filter depth, for example, or by providing dehumidifiers before the filter;

Providing shielding in places where radioactive material released to the containment or to a building would otherwise cause radiation exposure above the limits set for the accident analysis owing to direct or scattered radiation (including sky shine and ground shine);

Providing means of sealing the containment building or reducing the flow volume of exhaust air to provide for decay time within the building;

Reducing both the mass and activity of radioactive material released by decreasing the discharge velocity of fluids or the closure time of valves;

Ensuring the effectiveness of the spray system in trapping iodine by adding appropriate chemicals (e.g. hydrazine hydrate) or by adding chemicals in the reactor sump¹⁹ (e.g. sodium tetraborate).

Developing or upgrading safety systems and safety features for design extension conditions to minimize equipment malfunctions and operator errors, and to prevent and/or mitigate severe accidents;

Ensuring that power is available for essential equipment, instrumentation (including health physics instruments) and protection systems.

Provision of passageways of adequate dimensions for ease of access to plant systems and components. In areas where it is likely that site personnel will have to wear full protective clothing, including masks with portable air supplies or connections to a supply by air hoses, account should be taken of this in deciding on the dimensions of the passageways, noting that there may be an increase in the number of areas in which such equipment is necessary during decommissioning.

Provision of clear passageways of adequate dimensions to facilitate the removal of plant items to a workshop for decontamination or disposal. The routes for removing large items of plant during decommissioning should be planned at the design stage and the necessary provisions should be incorporated.

Facilitation of access to structures, systems and components, including compartmentalization of processes (e.g. through incorporation of hatches and large doors (see para. 7.6(b) of SSG‑47 [34])).

Provision of adequate space in the working areas to carry out decommissioning actions.

Provision of ladders, access platforms, crane rails or cranes in areas where their use is foreseen for the removal of plant components during decommissioning. Consideration should be given to access by robotic demolition machines as well as human access. Features to facilitate the installation of temporary shielding should be included in the design.

Use of computer aided design models to optimize decommissioning aspects of the design that affect working times during decommissioning. Video or photographic records should be made, and visualization modelling should be performed during the construction of the plant to facilitate the planning of work in areas of high radiation levels during decommissioning and thus to shorten working times.

Provision of means for the quick and easy removal of shielding and insulation during decommissioning.

Provision of special tools and equipment to facilitate work during decommissioning and thus reduce exposure times.

Provision of remote controlled equipment.

Provision of a suitable system for communication during decommissioning with the site personnel working in radiation areas or contamination areas.

Access controls for areas where dose rates and/or contamination can be temporarily high.

Provision of decontamination facilities and storage facilities with sufficient space for radioactive waste.

The use of modular construction to facilitate the dismantling of structures, systems and components;

Facilitation of the removal and/or decontamination of material and equipment, including by means of built‑in decontamination mechanisms, such as protective coverings and liners in process cells and areas where liquids might be present;

For materials that may be exposed to neutron radiation or materials in contact with reactor coolant, use of materials that are resistant to activation, that are resistant to degradation by chemicals and that have sufficient wear resistance to minimize the spread of activated corrosion products;

Consideration of provisions for the installation of ‘test specimens’ to facilitate the radiological characterization of structures, systems and components;

Measures (e.g. flushing) to avoid the sedimentation of radioactive sludge in piping and containers used during decommissioning;

Development of a waste management concept, especially concerning treatment of radioactive material for clearance or disposal, and options for logistics;

Consideration of provisions for water supply and drainage systems.

Protective clothing (e.g. coveralls, extra coveralls, helmets, eye protection, puncture resistant gloves, puncture resistant safety shoes, boots, shoe covers);

Respiratory protective equipment;

Air samplers and equipment for measuring airborne activity concentrations;

Portable dose rate meters with an alarm at variable settings and devices for monitoring personnel contamination and surface contamination;

Portable shielding, signs, ropes, stands and remote handling tools;

Communication equipment;

Meteorology instruments;

Equipment for monitoring individuals for intakes of radionuclides;

Temporary containers for solid radioactive waste and special containers for radioactive liquids;

Emergency equipment (including additional specialized protective clothing, air samplers with built‑in power sources and emergency vehicles);

First aid equipment;

Equipment for sampling and analysis around waste storage areas, such as borehole monitoring equipment for underground storage facilities for radioactive waste;

Transport and handling equipment provided in the nuclear power plant design (for the operating stage), for dismantling structures, systems and components and for waste management during decommissioning;

Personal dosimeters for monitoring individual external exposure.

Mechanisms of thermal and mechanical mixing;

The limited effectiveness of dilution in reducing airborne contamination;

Extraction of air from areas of potential contamination at points near the source of the contamination;

The use of exhaust rates that are commensurate with the potential for contamination in the area;

The need to ensure that the exhaust air discharge point is not close to an intake point of the ventilation system.

Maintaining the confinement of stored materials;

Providing for radiation protection (by means of shielding and contamination control);

Providing for ventilation, as necessary;

Allowing for retrieval and packaging of waste for transport off the site;

Further design features against incidents and accidents.

Generation of hazardous gases caused by chemical effects and the buildup of overpressure;

Generation of combustible or corrosive substances.

Area monitoring systems within the plant;

Individual monitoring of external exposure and internal exposure of workers;

Monitoring of discharges;

Environmental monitoring, including measurements of background radioactivity in the environment;

Monitoring of other parameters important for the assessment of public exposure (e.g. environmental and meteorological conditions at the site);

Process system monitoring systems within the plant.

Range of dose rates or activity concentrations to be measured;

Physical quantities and units displayed and archived;

Device sensitivity and accuracy;

Uncertainty of measurement results;

Types of radiation or radionuclides to be monitored;

Provision of threshold alarms;

Power supply and backup power supply;

Environmental conditions;

Provision for testing, calibration and maintenance;

Provision for functioning in all plant states, including accident conditions;

Response to overload conditions;

Failure mode indication;

Potential for interference with or corruption of monitored data due to other radiation sources present in the area, for example in the case of monitoring for neutron radiation and tritium;

Connectivity of the equipment for data transfer;

Provisions for saving and accessing results.

The reactor containment;

Rooms that are adjacent to the upper part(refuelling area) of the containment;

Buildings that are housing systems connected to the reactor coolant system;

The spent fuel storage facility;

The fuel handling machine;

The treatment and storage facilities for radioactive waste;

The decontamination facilities;

The transport routes for fuel and waste;

Areas for the handling and storage of fresh mixed oxide fuel or fresh fuel containing reprocessed uranium;

The main control room and supplementary control room;

Emergency response facilities;

Outdoor locations in and around the site.

The containment;

The fuel storage facilities;

The auxiliary building;

The radioactive waste building.

Equipment for monitoring contamination of skin and clothing;

Equipment for monitoring contamination of any objects or material being removed from the area.

Plant off‑gas system;

Vent header of radioactive waste tanks;

Building ventilation with the potential for contamination;

Liquid effluents of the radioactive waste system;

Condenser cooling water discharge.

Exposure pathways to people, including the food chain;

Radiological impacts, if any, on the local environment;

The possible buildup, and accumulation in the environment, of radioactive substances;

The possibility of any unauthorized routes for radioactive releases.”

Storage, cooling and cleanup systems for irradiated fuel;

Sumps connected to drain systems for radioactive liquids;

Ventilation ducts for radioactive discharges;

Circuits or systems separated by only one barrier from radioactive circuits (e.g. that might become contaminated owing to leaks in heat exchangers).

A considerable amount of data have been obtained from operating plants on parameters that are relevant to the exposure of site personnel and members of the public.

Progress has been made in understanding the phenomena that determine the production and transport of radioactive material within the plant.

Specialized computer software has been developed to make predictions for situations where the quality of the data is poor or where significant features of the design have been changed.

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INTERNATIONAL ATOMIC ENERGY AGENCY, Safety of Nuclear Power Plants: Commissioning and Operation and IAEA Safety Standards Series No. SSR‑2/2 (Rev. 1), IAEA, Vienna (2016). https://doi.org/10.61092/iaea.cq1k-j5z3

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The quantity of fissile material;

The fuel power and burn‑up;

The neutron flux distribution in the core;

The operational power history (including transients) and fuel management;

Decay time after reactor shutdown.

Reducing the corrosion and erosion rate by the proper selection of materials and the control of the coolant chemistry;

Selecting materials to minimize the concentration of nuclides (particularly of cobalt in steel) that are known to become major sources of radiation;

Providing removal systems (such as particulate filters and ion exchange resins);

Minimizing the concentration of nuclides in feedwater that can be activated in the core;

Providing surface treatments to reduce the adherence of corrosion products to pipes and components.

High purity (e.g. leakage wastes from the primary circuit of pressurized water reactors during power operation);

High chemical content (e.g. decontamination agents);

High solid content (e.g. liquid wastes from floor drains);

Liquid wastes containing detergent (e.g. liquid wastes from laundry drains and personnel showers);

Liquid wastes containing oil (e.g. in gas cooled reactors, liquid wastes from floor drains from the area of the lubricating oil tank for the circulator);

Liquid wastes with a very high tritium content (from pressurized heavy water reactors).

Components and structures that become activated or contaminated and have to be removed (e.g. control rods, neutron source assemblies, defective pumps, flux measuring assemblies, structures or parts thereof);

Irradiated components of the fuel assembly from gas cooled reactors (in which the assemblies are dismantled at the nuclear power plant);

Ion exchange resins, filter material, filter coating material, catalysts, desiccants and similar;

Concentrates from evaporators, precipitates;

Contaminated tools;

Contaminated clothing, towels, plastic sheet, paper and similar.

The calandria vault gas (in pressure tube reactors).

The cover gases of containers in which liquids with some content of volatile substances are stored (e.g.storage tanks for collected reactor coolant leakage water in light water reactors and storage tanks or some other equipment in the liquid waste treatment system). In some cases, gases are formed by decay (e.g. the decay of iodine to xenon).

Coolant gas leaking into sections that contain air in gas cooled reactors.

Air that has entered the pressure vessel of a light water reactor after it has been depressurized and the water level has been lowered prior to opening the vessel.

The high activity concentration for the leakage;

The break being not fully isolable;

The dry out of the affected steam generator that results in no partitioning of radioactive material within the steam generator.

Spent fuel handling areas (i.e. fuelling machines, the dry fuel store, the fuel dismantling cell, the fuel storage pool, and the loading bay for fuel transport flasks);

The active effluent treatment plant;

The treatment and cooling plant for fuel pool water;

The coolant treatment plant;

The storage area for solid radioactive waste;

The fuel element debris vault;

Ventilation filters.

Coolant activity release. During this phase, (relatively) small amounts of radioactivity in the coolant itself are released from the reactor coolant system.

Gap inventory release. This phase involves the release of radioactivity from the gap between the fuel pellets and cladding, including noble gases, iodine and caesium.

In‑vessel release. During this phase, which is associated with gradual melting and slumping of core materials, practically all of the noble gases and significant fractions of the volatile nuclides are released from the fuel.

Ex‑vessel release. If failure of the reactor pressure vessel occurs, the volatile radionuclides not released during the in‑vessel phase are released from the molten corium into the containment. In existing designs, the molten core interacts with the reactor cavity concrete, and the less volatile nuclides are likely to be released into the containment. The presence of water in the reactor cavity overlying core debris can significantly reduce the ex‑vessel releases either by cooling the core debris or by scrubbing the releases and retaining a large fraction in the water.

Late in‑vessel release. Simultaneously with ex‑vessel release, some of the volatile nuclides that had deposited in the reactor coolant system during the in‑vessel phase are released into the containment.

Parameters relating to the solubility in water of unsaturated nickel, nickel oxide, nickel ferrites and cobalt ferrites at a coolant temperature range of 280°C–340°C and a pH range of 6.5–7.4 at 300°C are very important for determining the behaviour of corrosion products in the primary coolant.

The models that are used to describe the behaviour of corrosion products need to have the capability to model a large interacting ‘water–metal’ system for which the following parameters are typical:

— Area in contact with the primary coolant: ~30 000 m²;

— Mass of coolant: 200–300 t;

— Velocity of coolant: 0.1–15 m·s^–1.

The duration of one circuit (reactor → steam generator → reactor): ~10 s including ~1 s in flux.

The variety of alloys: Zircaloy 4/Inconel 600, Inconel 690, Incoloy 800/ Inconel 718/hardfacing materials (Stellite)/stainless steel.

The order of magnitude of the mass of precursors of radioactive species (essentially ⁵⁸Ni (n, p) ⁵⁸Co and ⁵⁹Co (n, g) ⁶⁰Co) is as follows:

— Release (average): 1 mg·dm^-2 ·month^-1 ;

— Cycle duration: 12–18 months;

— Area excluding Zircaloy (~ no release): 17 000 m² ;

— ⁵⁹Co level (impurity): ~5·10^–4 g·g^–1;

— ⁵⁸Ni in nickel‑based alloys (Inconel 600, 690): ~3·10^–1 g·g^–1.

The input to the reactor coolant during a ten month cycle is ~10 g·cycle–1 of ⁵⁹Co and ~5 kg·cycle–1 of ⁵⁸Ni.

Wear of hardfacing materials (e.g. of parts in the internal structures of the core, pump bearings, valves, control rod drive mechanisms) is in addition to the figure for ⁵⁹Co.

As a result, approximately 10 g of ⁵⁹Co and 5 kg of ⁵⁸Ni are the origin of the ⁶⁰Co and ⁵⁸Co deposits, respectively, that are responsible for 90% of the dose rates and occupational exposures. This applies for reactors in which a nickel‑based alloy is used for steam generator tubing.

Ionic species can precipitate and agglomerate to particles. These particles circulate in the fluid and are likely to form deposits either within the reactor core or on out‑of‑flux surfaces. By this process they become activated during circulation or after they have been deposited on in‑core surfaces.

Ions and particles can be removed from the primary coolant by the coolant purification system. The effectiveness of this process depends on the flow rate and on the decontamination factors of the filters and the ion exchange columns of the coolant purification system. If any of these factors are too low, the purification system will be ineffective.

When the concentration of oxides in the primary coolant is very low (in a pressurized water reactor it is typically a few 10^–9 g·g^–1);

When the release of elements from alloys is not proportional to their composition;

When chemical conditions vary throughout the fuel cycle within a specified range;

When bulk coolant and surface temperatures need to be taken into account;

When wear by friction is significant.

Calculating the inventory of fission products in the fuel — several well‑known computer codes are available to perform this evaluation (e.g. the ORIGEN, FISPIN, APOLLO and THALES2 codes in the United States of America, United Kingdom, France and Japan, respectively).

Determining the activity of radionuclides in the gas gaps in the fuel elements.

Determining the total activity of the radionuclides released to the coolant through cladding defects.

Determining the transport of the radionuclides in the reactor coolant system (primary, secondary and auxiliary systems, as applicable) and the containment, up to the determination of the source term to the environment.