040.1411.131112
ՀԱՅԱՍՏԱՆԻ ՀԱՆՐԱՊԵՏՈՒԹՅԱՆ ԿԱՌԱՎԱՐՈՒԹՅՈՒՆ
ՈՐՈՇՈՒՄ
8 նոյեմբերի 2012 թվականի N 1411-Ն
ՄԻՋՈՒԿԱՅԻՆ ՆՈՐ ԷՆԵՐԳԱԲԼՈԿԻ (ԷՆԵՐԳԱԲԼՈԿՆԵՐԻ) ՆԱԽԱԳԾԱՅԻՆ ԱՆՎՏԱՆԳՈՒԹՅԱՆ ՊԱՀԱՆՋՆԵՐԸ ՀԱՍՏԱՏԵԼՈՒ ՄԱՍԻՆ
(2-րդ մաս)
1. no single failure results in loss of protection function; and
2. the removal from service of any component or channel does not result in loss of the necessary minimum redundancy.
The protective safety systems shall be designed to ensure that the effects of normal operation, anticipated operational occurrences and design basis accidents on redundant channels do not result in loss of function; unless such a loss shall be demonstrated to be acceptable on some other basis.
All reactor designs shall include safety features to mitigate the consequences of an anticipated transient without a reactor scram (ATWS).
Supporting safety systems. Auxiliary services that support equipment forming part of protective safety systems or other systems important to safety shall be classified according to the significance of the systems they support. Their reliability, redundancy, diversity and independence and the provision of features for isolation and for testing of functional capability shall be commensurate with the reliability of the system that is supported.
Auxiliary services necessary to maintain the plant in a safe state may include the supply of electricity, cooling water and compressed air or other gases, and means of lubrication.
Systems shall be provided to transfer residual heat from SSC important to safety to an ultimate heat sink. This function shall be carried out at very high levels of reliability in all operational states and in design basis accidents. All such systems shall be designed in accordance with the importance of their contribution to the function of heat transfer.
The reliability of the systems shall be achieved by the use of proven components, redundancy, diversity, physical separation, interconnection and isolation.
Design basis natural phenomena and human induced events shall be taken into account in the design of the systems; and in the possible choice of diversity in the ultimate heat sinks; and in the storage systems from which fluids for heat transfer are supplied.
Adequate consideration shall be given to extending the capability to transfer residual heat from the core to an ultimate heat sink so as to ensure that, in the event of a severe accident, acceptable temperatures can be maintained in structures, systems and components important to the safety function of confinement of radioactive materials.
Equipment qualification. A qualification program shall be adopted to confirm that SSC important to safety can perform their functions throughout their design operational lives while being subjected to the environmental conditions (of vibration, temperature, pressure, jet impingement, electromagnetic interference, irradiation, humidity or any likely combination thereof) prevailing at the time of need.
The environmental conditions to be considered shall include the variations expected in normal operation, anticipated operational occurrences and design basis accidents.
Consideration shall be given to aging effects caused by environmental factors (such as vibration, irradiation and extreme temperature) over the expected lifetime of the equipment. Where the equipment is subject to external natural events and is needed to perform a safety function in or following such an event, the qualification program shall replicate as far as practicable the conditions imposed on the equipment, either by test or by analysis or by a combination of both.
In addition, any unusual environmental conditions that can reasonably be anticipated and that could arise from specific operational states, such as in periodic testing of the containment leak rate, shall be included in the qualification program. To the extent possible, equipment (such as certain instrumentation) that must operate during a severe accident should be shown, with reasonable confidence, to be capable of achieving the design intent.
Protection from ??F. The NPP design shall consider and justify measures for prevention or protection of systems and components from common-cause failures. The potential for common cause failures of systems and components important to safety shall be considered to determine where the principles of diversity, redundancy and independence should be applied to achieve the necessary reliability.
Human engineering. The design shall be «operator friendly' and shall be aimed at limiting the effects of human errors. Attention shall be paid to plant layout and procedures (administrative, operational and emergency), including maintenance and inspection, in order to facilitate the interface between the operating personnel and the plant. The working areas and working environment of the site personnel shall be designed according to ergonomic principles.
Consideration of human factors and the human-machine interface shall be included at an early stage and throughout the entire design process, to ensure an appropriate and clear distinction of functions between operating personnel and the automatic systems is identified. The human-machine interface shall be designed to provide the operators with comprehensive but easily manageable information, compatible with the necessary decision and action times. Similar provisions shall be made for the supplementary control room.
Verification and validation of aspects of human factors shall be included at appropriate stages to confirm that the design adequately accommodates all necessary operator actions.
The operator shall be considered to have dual roles: that of a systems manager, including accident management, and that of an equipment operator. In the systems manager role, the operator shall be provided with information that permits the following:
1. the ready assessment of the general state of the plant, whether in normal operation, during an anticipated operational occurrence or in an accident condition, and confirmation that the designed automatic safety actions are being carried out; and
2. the determination of the appropriate operator initiated safety actions to be taken.
As equipment operator, the operator shall be provided with sufficient information on parameters associated with individual plant systems and equipment to confirm that the necessary safety actions can be initiated safely.
Operator actions necessary for safe operation shall account for the time available for action, the physical environment to be expected and the psychological demands on the operator. The need for action on a short time-scale shall be kept to a minimum.
The design shall include means for prevention of single operator errors or mitigation of their consequences, including those during maintenance.
Avoiding multiple use of systems. Each unit shall have its own systems important to safety to control and mitigate the anticipated operational occurrences and accidents considered for the design. Safety systems shall not be shared between multiple units unless this contributes to enhanced safety.
In accident conditions, inter connecting support systems among the units is allowed if it can be justified that it facilitates the accident management of one unit by giving the possibility to restore a safety function. Safety system support features and safety related items shall be permitted to be shared between several units of a nuclear power plant if this contributes to safety. Such a sharing shall not be permitted if it would increase either the likelihood or the consequences of an accident at any unit of the plant.
System interaction. The potential for harmful interactions of systems important to safety at the nuclear power plant that might be required to operate simultaneously shall be evaluated, and effects of any harmful interactions shall be prevented. In the analysis of the potential for harmful interactions of systems important to safety, due account shall be taken of physical interconnections and of the possible effects of one system's operation, maloperation or malfunction on local environmental conditions of other essential systems, to ensure that changes in environmental conditions do not affect the reliability of systems or components in functioning as intended. If two fluid systems important to safety are interconnected and are operating at different pressures, either the systems shall both be designed to withstand the higher pressure, or provision shall be made to prevent the design pressure of the system operating at the lower pressure from being exceeded.
Design to enable ISI/IST. SSC important to safety shall be designed to be calibrated, tested, maintained, repaired or replaced, inspected and monitored with respect to their functional capability over the lifetime of the plant to demonstrate that reliability targets are being met. The plant layout shall be such that access is provided for in-service inspection and in-service testing without undue exposure of the site personnel to radiation.
For safety-significant systems and components a direct and complete inspection shall be performed during the plant commissioning for compliance with design characteristics. Additionally, such inspections shall be performed periodically and after maintenance of these systems over the whole plant life-time period:
1. the design shall provide for possibility of diagnostics (tests) of safety systems and components belonging to classes 1 and 2, and possibility for their testing in conditions simulating an emergency situation to a maximum extent possible; and
2. frequency and allowed maintenance and testing time shall be justified in the design or approved according to a special procedure.
Design to enable decommissioning. The design shall incorporation features that will facilitate the decommissioning and dismantling of the plant. In particular, account shall be taken of:
* the choice of materials, such that eventual quantities of radioactive waste are minimized and decontamination is facilitated;
* the access capabilities that may be necessary; and
* the facilities necessary for managing of radioactive waste that will be generated in the decommissioning of the plant.
Physical Plant Security and Safeguards. The licensee shall provide physical protection against radiological sabotage and against theft of special nuclear material. The licensee shall establish and maintain physical security in accordance with security plans approved by the regulatory agency. The scope of the program shall include:
* protection against radiological sabotage, including determined violent external assault, attack by stealth, or deceptive actions;
* protection against theft or diversion of strategic special nuclear material;
* protection of spent nuclear fuel and high-level radioactive waste; and
* protection of physical security information.
The physical protection system shall include provisions to:
* prevent unauthorized access of persons, vehicles and materials into material access areas and vital areas;
* permit only authorized activities and conditions within protected areas, material access areas, and vital areas;
* permit only authorized placement and movement of strategic special nuclear material within material access areas;
* permit removal of only authorized and confirmed forms and amounts of strategic special nuclear material from material access areas;
* limit authorized access and assure detection of and response to unauthorized penetrations of the protected area, and
* limit location of vital equipment only within vital areas, and storage of strategic special nuclear material only in a material access area.
The plant security system shall include provisions for:
* a security organization, including guards;
* access control subsystems and procedures;
* detection, surveillance and alarm subsystems and procedures; and
* contingency and response plans and procedures for responding to security emergencies.
Design for ease of egress in emergency. The nuclear power plant shall be provided with a sufficient number of safe escape routes, clearly and durably marked, with reliable emergency lighting, ventilation and other building services essential to the safe use of these routes. The escape routes shall meet the relevant international requirements for radiation zoning and fire protection and the relevant national requirements for industrial safety and plant security.
Alarm systems and means of communication shall be provided so that all persons present in the plant and on the site can be warned and instructed, even under accident conditions.
The availability of means of communication necessary for safety, within the nuclear power plant, in the immediate vicinity and to off-site agencies, as stipulated in the emergency plan, shall be ensured at all times. This requirement shall be taken into account in the design and the diversity of the methods of communication selected.
Preference for passive systems. Safety systems shall preferably rely on passive devices and the fail-safe design principle (safe geometry, safe parameters, self-control, temperature difference and natural processes).
Accounting for external effects in design. The plant design shall consider site-specific man-induced and naturally caused external events, as well as geotechnical characteristics of foundation materials. The design basis shall use site-specific parameter values to characterize external hazards.
Man-induced external hazards with probability of occurrence below 10-7 events per year may be excluded from the design basis.
The assessment and design of protection from natural and man-induced external events shall consider loads due to the external hazards in combination with normal operating and transient loads.
Regardless of low levels of external impact intensity assumed in design basis, the design shall have the following provisions:
* seismic resistance to horizontal peak ground acceleration more than 0.4g. Seismic hazard at the site shall be defined on a basis of earthquake sources, determined site and site-specific response spectra, performance-based ground motion response spectra.
* the wind used in the design shall be the most severe wind that has been historically reported for the site and surrounding area with sufficient margin for the limited accuracy, quantity, and period of time in which historical data have been accumulated.)
* resistance to loads produced by explosive shock waves of not less than 10 kPa;
* resistance of safety-related buildings and structures to external fires-not less than 2.5 hours under an external thermal environment of up to 300oC;
* spatial and physical separation of safety systems and their trains;
* resistance of protective structures for confining systems to impact loads from a commercial airliner during normal landing. Design shall include section as follows:
1. missile effects on plant structures from aircraft impacts;
2. fire effects from aircraft fires;
3. requirements to protect plant SSCs important to safety from aircraft crashes.
SSC included in seismic Category I shall perform their safety functions during and after a seismic event, assuming ground motion equivalent to that of a safe shutdown earthquake (SSE).
Plant personnel shall be provided with protection from external factors. External loads on personnel shall be maintained within limits which do not degrade reliability (or welfare) of personnel.
The NPP design shall provide external event warning systems; recording of natural and man-induced external impacts; to determine if the maximum calculated level established by design basis is exceeded.
In DEC it shall be demonstrated that SSC (including mobile equipment and their connecting points, if applicable) for the prevention of fuel damage or mitigation of consequences in DEC have the capacity and capability and are adequately qualified to perform their relevant functions for the appropriate period of time.
6.1 Reactor Core and Associated Features
Reactor core and associated reactor coolant system, reactor control and protection safety systems shall be designed with appropriate safety margins to ensure that the specified acceptable design limits for fuel damage are not exceeded during all operational states and design basis accidents with account taken of:
* design operating modes and their passing;
* thermal, mechanical and irradiation degradation of the core components;
* physical-chemical interaction of core materials;
* limiting values of thermal hydraulic parameters;
* vibrations and thermal cycles, material fatigue and aging;
* impact of coolant additives and radioactive fission products on the corrosion of fuel cladding;
* irradiation and other impacts that deteriorate mechanical characteristics of core materials and fuel cladding integrity.
The design of the reactor core shall specify the limits for damage of fuel elements (in terms of amount and degree) and the associated coolant radioactivity according to reference isotopes.
To ensure safe shutdown of the reactor, to maintain the reactor subcritical and to ensure adequate core cooling, the reactor core and associated internal components located within the reactor vessel shall be designed and mounted in such a way as to withstand the static and dynamic loads expected in all operational states and external events considered in the design.
Reactor core and its elements that affect reactivity shall be designed in a way that any reactivity change caused by the control rods as well as reactivity effects shall not lead to fuel damage that exceeds the specified design limits and shall not cause any damage to reactor coolant pressure boundary during all operational states and design basis accidents.
Design shall be such that in all design basis accidents with fast insertion of positive reactivity, specific energy threshold for fuel damage is not exceeded at any moment of the fuel cycle and fuel melting is excluded by insertion of the control rods. With respect to beyond design basis accidents, conditions for possible fuel melting or exceeding the specific energy threshold causing fuel damage shall be specified.
For all design basis accidents and for beyond design basis accidents changes in core geometry shall be limited thus ensuring conditions for long-term fuel cooling.
The combined reactivity coefficients of coolant density, of coolant-moderator and fuel temperature, and of reactor power, shall be negative within the whole range of the reactor coolant system parameters for all operational states and design basis accidents.
Design shall ensure minimization of possibilities for re-criticality and reactivity excursions following postulated initiating events.
Design of the reactor core shall reduce demands on the system for control of the neutron flux (distribution, levels and stability within specified limits) in all operational states.
Reactor core and associated coolant, control and protection systems shall be designed to enable adequate inspection and testing throughout the service lifetime of the plant.
The characteristics of nuclear fuel, the reactor structures and of the reactor coolant system components (including the coolant clean up system) shall prevent re-criticality in severe accidents, considering the operation of the other systems.
Fuel elements and assemblies, taking into account the uncertainties in data, calculations and fabrication, shall be designed to withstand irradiation and reactor core conditions in combination with all degradation processes that can occur in all operation states, such as:
* differential expansion and deformation;
* external pressure of the coolant;
* additional internal pressure due to fission products in the fuel element;
* irradiation of fuel and other materials in the fuel assembly;
* changes in pressures and temperatures resulting from changes in power;
* chemical effects;
* static and dynamic loads, including flow induced vibrations and mechanical vibrations;
* changes in heat transfer that may be a result of distortions or chemical effects.
The reactor core and associated coolant, control and protection systems shall be designed with safety margins to ensure that the specified acceptable fuel design limits are not exceeded.
6.2 Reactor Coolant System
The design of reactor coolant system shall include passive pressure relieving devices (safety valves) with sufficient relieving capacity to prevent exceeding the reactor coolant design pressure and will not lead to the release of radioactive material from the nuclear power plant directly to the environment during all conditions of normal operation, abnormal operational occurrences and postulated accidents.
Components, pipelines and supporting structures of the reactor coolant system shall withstand all anticipated static and dynamic loads and temperature effects on the components during all postulated initiating events and external events such as seismic events.
Materials to be used for fabrication of the components of the reactor coolant system shall be selected so as to minimize the probability of crack propagation and neutron embrittlement, with account taken of the expected degradation of their characteristics at the end-of lifetime under the effects of erosion, creep, fatigue and chemical environment.
Reactor pressure vessel shall be designed and constructed to be of the highest quality with respect to material selection, design standards, capability of inspection and fabrication.
Design of the components contained inside the reactor coolant pressure boundary shall be such as to minimize the likelihood of failure and associated consequential damage to other items of the primary coolant system in all operational states and in design basis accidents.
Components of the reactor coolant pressure boundary shall be designed, manufactured and situated in a way allowing periodical inspections and tests to be carried out, throughout the service lifetime of the plant. Implementation of a material surveillance program for the reactor coolant pressure boundary shall monitor the effects on structural materials of various factors such as irradiation, stress corrosion cracking, embrittlement, and ageing and particularly in locations of high irradiation, and others.
Provisions shall be made in the design to regulate coolant inventory and pressure with adequate capacity for all operational states.
Design shall provide for systems to cleanup reactor coolant from radioactive substances, including activated corrosion products and fission products. Capacity of the necessary systems shall be based on the fuel design limits on permissible leakage with a conservative margin to ensure that the coolant activity is as low as reasonably practical and that sub-criticality is assured following anticipated operational occurrences and during accidents.
The reactor coolant system shall be designed to prevent the initiation of flaws (cracks) at its pressure boundary. If initiation were to occur, the design shall be such that flaws will propagate in a metallurgical regime characterized by high resistance to unstable fracture and rapid crack propagation. Designs and plant states in which components of the reactor coolant pressure boundary could exhibit brittle behavior shall be avoided.
The design of components inside the reactor coolant pressure boundary, such as pump impellers and valve parts, shall minimize the likelihood of failure and consequential damage to other items of the primary coolant system in all operational states and in design basis accidents, with due allowance for deterioration that may occur in service.
6.3 Removal of Residual Heat
Plant design shall provide for redundant safety-related systems to remove, to an ultimate heat sink, the residual heat from the core and from SSCs important to safety, in all operational states and design basis accidents. All systems that contribute to the heat transfer (by conveying heat, by providing power or by supplying fluids to the heat transport systems) shall be designed and classified according to their safety significance. Systems interfacing directly with the reactor coolant system shall be Class 2. The portion of the RHR system that is not isolable from the RCS shall be classified as Class 1. Supporting safety-related systems such as component cooling water and service water shall be classified as Class 3.
Reliability of the systems shall be achieved by the use of proven components, redundancy, diversity, physical separation and isolation.
Natural phenomena and human induced events specific to the NPP site shall be taken into account in the design of the systems and in the possible choice of diversity in the ultimate heat sinks.
Adequate consideration shall be given to the residual heat removal from the reactor core and cooling of the localization system components in case of a severe accident.
6.4 Emergency Core Cooling
Core cooling shall be provided in the event of a loss of coolant accident so as to minimize fuel damage and limit the escape of fission products from the fuel. The cooling provided shall ensure that:
1. limiting parameters for cladding or fuel integrity (such as temperature) will not exceed acceptable values for design basis accidents;
2. possible chemical reactions are limited to an allowable level;
3. the alterations in the fuel and internal structural alterations will not significantly reduce the effectiveness of the means of emergency core cooling, and
4. the cooling of the core will be ensured for a sufficient time.
Adequate consideration shall be given to extending the capability to remove heat from the core following a severe accident.
6.5 Heat transfer to an ultimate heat sink
Systems shall be provided to transfer residual heat from items important to safety at the nuclear power plant to an ultimate heat sink. This function shall be carried out with very high levels of reliability for all plant states.
If a high level of reliability of the residual heat transfer to the ultimate heat sink cannot be demonstrated for all potential conditions generated by external hazards, alternative means shall be provided. These means, including the use of a different heat sink and the necessary associated features, shall be located and designed so that external hazards cannot result in the loss of the residual heat removal function.
6.6 Control of the Technological Processes
NPP unit shall be provided with the following means to control and monitor the systems for normal operation and the safety systems:
1. main control room (MCR);
2. supplementary control room (SCR);
3. control systems for systems required during plant shutdown, refueling and normal operation;
4. control systems for safety-related engineered safety features required for accident mitigation;
5. independent means for information collection and storage.
The MCR shall provide possibilities for undertaking measures to maintain the plant in a safe state or to recover such state if needed in all operational states and design basis accidents.
Design shall be sufficient to maintain the MCR personnel health and availability, as well as proper functioning of the MCR, in all operational states and internal and external events.
MCR design shall provide for:
1. instrumentation to control the fission process, in all core states, and conditions in normal operation, including in subcritical state during refueling;
2. position indicators of the reactivity control devices, automatic control of soluble neutron absorber concentration, and status indicators of all means for reactivity control;
3. a system for information support to the operators;
4. a safety parameter display system of the reactor installation.
Control signals of technological systems and components important to safety, formed by the automatic control system or by the MCR remote control switches, shall be automatically registered.
Any possibility of parallel actuation of control components, from the MCR and the SCR, shall be eliminated by technical means. Appropriate measures shall be taken to eliminate any possibility for failure of the control circuits of both MCR and SCR due to a common cause, in all postulated initiating events.
SCR shall be designed to protect the personnel in all conditions resulting from internal and external events and design basis accidents.
Control systems for normal operation shall control and regulate the technological processes, in all operational states, in conformity with the design specified indicators for quality, reliability and metrological characteristics, and shall encompass:
1. means for collecting, treating, documenting and storing of information, which to be sufficient for timely and unambiguous identification of the initiating events for anticipated operational occurrences and accidents, their progression, factual algorithms of operation of the safety systems and the components, which failures are initiating events for design basis and beyond design basis accidents, deviations from the design algorithms and personnel actions;
2. means for automatic control of reactor coolant activity, liquid and gaseous effluents to the environment, and radiation monitoring of plant compartments, and radiation protection and monitored areas, in all operational states and design basis accidents;
3. means for automatic control of the conditions for safe storing of nuclear fuel and radioactive waste, and for notification in case these conditions are violated;
4. means and methods for identification of the locations and quantities of coolant leakages;
5. means for reliable group and individual communications between the MCR, SCR and field operators.
Control systems for normal operation shall ensure the most favorable conditions to the operating personnel to take the correct decisions for plant management.
In the design of computer based control systems for normal operation:
1. special standards and proven practices shall be used in development and verification of the hardware, and especially of the software;
2. development and verification process shall be conducted in compliance with a quality assurance program;
3. level of reliability assumed in the safety analysis shall include a specified conservatism to compensate for the inherent complexity of the technology.
Control safety systems shall be designed to:
1. initiate automatically the operation of appropriate systems, including systems for reactor shutdown, in order to ensure that specified design limits are not exceeded as a result of anticipated operational occurrences;
2. detect the symptoms for design basis accidents and automatically actuate other safety systems necessary to limit the consequences within the design basis;
3. lock the switch-off capability of the operating personnel for at least 30 minutes after an automatic actuation;
4. be capable of overriding unsafe actions of the control systems for normal operation.
Design shall provide possibilities for manual remote actuation of the safety systems - and for the isolation of components at their location. A failure in automatic actuation circuits shall not impede the remote manual actuation and the implementation of the safety functions.
Design of control safety systems shall provide for: continuous automatic diagnostics of the systems operability; periodic testing from MCR and SCR of system channels; and diagnosis of the technological components,
Instrumentation shall be provided to monitor plant variables and the status of essential equipment over the ranges for normal operation, anticipated operational occurrences, design basis accidents and severe accidents; to allow projections of the locations and quantities of radioactive materials that could escape from the plant; and to permit classifying events for the purposes of emergency response. Instrumentation and recording equipment shall be adequate for determining plant status in a severe accident and for making accident management decisions.
6.7 Containment System
Reactor installation design shall include containment safety systems to ensure fulfillment of the established criteria for radioactive releases to the environment. Containment safety systems shall perform their functions in all postulated initiating events and mitigate the consequences of beyond-design basis accidents including severe accidents.
In establishment of containment functions, provisions shall include a leak tight structure, systems and means for control of containment parameters, for containment structure isolation, and for reducing the concentration of fission products, hydrogen and other substances that could be released in the containment atmosphere during and after design basis and severe accidents.
The containment structure and its components, including hermetic access doors, penetrations and isolation devices, shall be designed with sufficient safety margins on the basis of potential internal overpressure, underpressure and temperatures, dynamic effects such as missiles impact, reaction forces, and the effects of other potential energy sources anticipated to arise as a result of design basis accidents.
In calculating the necessary strength of the containment structure and its components, natural phenomena and human induced events shall be taken into consideration, as well as a combination of the effects of reactor coolant system break with maximum size and safe shutdown earthquake.
The containment structure and its components shall be designed and constructed to permit structural integrity testing during commissioning and performing of periodic leaktightness tests over plant lifetime. The design shall specify test requirements and the respective methods and means. Components located inside the containment shall retain their functional capability after the tests have been conducted.
Number of penetrations through the containment structure shall be kept to a practical minimum. All penetrations shall meet containment structure design requirements with account of possible mechanical, thermal, and chemical effects.
Elastic components of containment penetrations shall be designed to allow individual leak testing, independent of the containment leak rate detection (integral test).
To prevent radioactive releases outside the containment in case of a design basis accident, any containment penetrating line (part of the reactor coolant pressure boundary or directly connected to containment atmosphere) shall be reliably isolated by at least two isolation valves having independent automatic control, arranged in series and located outside and inside the containment structure as close to the containment structure as practicable.
Any containment penetrating line that is neither directly connected to the reactor coolant pressure boundary nor to the containment atmosphere shall be reliably isolated by at least one isolation valve outside the containment and located as close to the containment structure as practicable.
To secure personnel access to containment premises, provisions shall be made of lock and block doors as to secure at least one door in a locked position for all operational states and design basis accidents.
The design shall include arrangements to ensure capability of isolation devices to maintain their functionality in the event of a severe accident.
Containment design shall include measures and technical means to ensure sufficiently low pressure difference between the separate internal compartments so as not to challenge the integrity of pressure bearing structure or of other systems with containment functions, taking into account the pressure and the possible effects resulting from design basis accidents.
The capability to remove heat from the containment shall be ensured, in order to reduce the pressure and temperature in the containment, and to maintain them at acceptably low levels after any accidental release of high energy fluids. The systems performing the function of removal of heat from the containment shall have sufficient reliability and redundancy to ensure that this function can be fulfilled.
The design shall include the necessary features to enable the use of non- permanent equipment to restore the containment cooling. The non-permanent equipment may be available at the site or not.
The loss of containment structural integrity shall be practically eliminated. This shall be achieved without significant radioactive releases.
Design features to control fission products, hydrogen, oxygen and other substances that might be released into the containment shall be provided as necessary:
* To reduce the amounts of fission products that could be released to the environment in accident conditions;
* To control the concentrations of hydrogen, oxygen and other substances in the containment atmosphere in accident conditions so as to prevent deflagration or detonation loads that could challenge the integrity of the containment.
Coverings, thermal insulations and coatings for components and structures within the containment system shall be carefully selected and methods for their application shall be specified to ensure the fulfillment of their safety functions and to minimize interference with other safety functions in the event of deterioration of the coverings, thermal insulations and coatings.
6.8 Emergency Power Supply
The emergency power supply at the nuclear power plant shall be capable of supplying the necessary power in anticipated operational occurrences and in design basis accidents in the event of the loss of off-site power. The design shall also include a dedicated power source to supply the necessary power in design extension conditions.
The design shall provide for emergency power supply system maintenance, periodic testing, tests and inspection of individual components, parts and trains during the whole service life in the process of operation and after maintenance.
6.9 Interactions between the electrical power grid and the plant
Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained. An exception to the requirement for two circuits may be made for a reactor whose safety systems do not rely on offsite power.
The trip of a nuclear power plant can affect the grid so as to result in a loss of offsite power. Foremost among such effects is a reduction in the plant's switchyard voltage as a result of the loss of the reactive power. Less likely results of the trip of a nuclear plant are grid instability, potential grid collapse, and subsequent loss of offsite power due to the loss of the real and/or reactive power support supplied to the grid from the plant's generator. To mitigate these events provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies. The offsite power circuits shall be designed to be available following a trip of the nuclear power unit(s), to permit the functioning of system structures and components necessary to respond to the event.
Procedures should include the actions necessary to restore offsite power and use nearby power sources when offsite power is unavailable. As a minimum, the following potential causes for loss of offsite power should be considered: grid under-voltage and collapse, weather-induced power loss, and preferred power distribution system faults (including distribution system hardware, switching and maintenance errors, and lightning-induced faults) that could result in the loss of normal power to essential switchgear buses.
6.10 Supporting and Auxiliary Systems
The design of supporting systems and auxiliary systems shall be such as to ensure that the performance of these systems is consistent with the safety significance of the system or component that they serve at the nuclear power plant.
Systems' design shall provide possibility for testing of their functional capability and for failure indication.
Fulfillment of supporting functions shall have priority over supporting systems own protections, if this will not aggravate safety consequences. Design shall specify the non-isolable own protections of the components of the supporting safety systems.
Design shall make provisions for fire alarm and fire-extinguishing systems to prevent fire-induced common cause failures in safety systems and to automatically fulfill the specified functions.
Fire-extinguishing systems shall also be able to be manually actuated.
7. RADIOACTIVE WASTE MANAGEMENT
Nuclear power plant design shall include systems and facilities for the management of radioactive waste generated in operation and in the decommissioning of the plant.
Radioactive waste (RAW) management systems shall be designed based on analysis and assessment of the composition and quantities of solid and liquid RAW and the gaseous radioactive substances generated in all operational states.
Systems for management of liquid and gaseous radioactive release to the environment shall be designed so that their quantities and concentrations are kept as low as reasonably achievable in all operational states and within the specified dose limits for the personnel and the population.
NPP design shall include systems for handling, treatment and storage of RAW in a condition suitable for transportation, further treatment and disposal.
NPP design shall include facilities for storage of RAW, equipped with remote means for manipulation.
RAW storage compartments shall be watertight and provided with systems for ventilation, decontamination, fire-alarm and fire-extinguishing.
The cleanup equipment for the gaseous radioactive substances shall provide the necessary retention factor to keep radioactive releases below the authorized limits on discharges. Filter systems shall be designed so that their efficiency can be tested, their performance and function can be regularly monitored over their service life, and filter cartridges can be replaced while maintaining the throughput of air.
Systems for treatment and control of the waste shall be provided at the nuclear power plant to keep the amounts and concentrations of radioactive releases below the authorized limits on discharges and as low as reasonably achievable.
Systems for the treatment and control of effluents Systems shall be provided at the nuclear power plant for treating liquid and gaseous radioactive effluents to keep their amounts below the authorized limits, as well as exposure of members of the public due to discharges to the environment is as low as reasonably achievable.
8. FUEL HANDLING
SSCs for handling and storage of non-irradiated fuel shall be designed to:
1. prevent criticality by a sufficient margin specified in other national regulations, even under the most adverse states, by ensuring related physical means or processes, such as geometrically safe configurations, and characteristics of the components and medium;
2. permit appropriate fuel acceptance test, maintenance, periodic inspection and testing of components important to safety;
3. ensure control of the storage conditions;
4. minimize the possibility of damage or unauthorized access to nuclear fuel;
5. prevent fuel assembly drop during transportation;
6. prevent the inadvertent dropping of heavy objects upon the fuel assemblies.
SSCs for handling and storage of irradiated fuel shall be designed in compliance with the requirements to non-irradiated fuel and additionally shall have the following:
1. reliable systems for residual heat removal during all operational states and design basis accidents;
2. measures to prevent unacceptable handling stresses on the fuel assemblies;
3. means for safe storage of non-tight or damaged fuel assemblies or fuel elements;
4. systems for local ventilation and other means for radiation protection;
5. means for identification of the fuel assemblies.
For reactors using a water pool system for storage of irradiated fuel, the design shall provide for the following:
1. means to control the temperature, water chemistry and activity;
2. means to monitor and control the water level in the storage pool and to detect leakages;
3. measures to prevent emptying the pool and uncovering of fuel assemblies as a result of syphon effect in the event of a pipe break;
4. means to control the concentration of the soluble neutron absorber.
Capacity of the structures for storing of irradiated fuel shall be substantiated in the design considering the capability at any time to completely remove the fuel from the reactor core.
9. RADIATION PROTECTION
During design the principles of radiation protection specified in Radiation Safety Norms of RA (Government Decree No1498-N as of 18.08.2006) shall be applied.
To ensure radiation protection, NPP design shall identify all real and potential sources of ionizing radiation and shall provide measures for ensuring the technical and administrative control over their use according to the national legislation.
To keep the exposure of personnel and public as low as reasonably achievable during plant operation, the design of the reactor coolant system shall arrange for:
1. use of structural materials with minimum content of chemical elements with high activation cross-section and producing long-living radioactive corrosion products;
2. coolant purification from fission and corrosion products;
3. water chemistry control;
4. minimum length of the pipelines with a minimum number of isolation valves and connections;
5. leak-tightness testing of operating components;
6. decontamination of SSCs outer and inner surfaces;
7. prevention of uncontrolled radioactive leaks in the NPP premises.
The layout of the plant, its buildings and SSCs shall facilitate the operation, inspections, maintenance, refueling, repair and replacement of systems and components and shall limit the personnel exposure to ionizing radiation.
The buildings, compartments and components, which may be contaminated with radioactive substances, shall be designed in a way that allows easy decontamination by chemical or mechanical means.
Facilities shall be provided for the decontamination of operating personnel and plant equipment which may be contaminated.
Plant equipment subject to frequent maintenance or manual operation shall be located in areas of low dose rate to reduce the exposure of workers.
Systems and system components shall be designed in a way that allows for the operation of the nuclear power plant without being present or working at high radiation areas.
Remotely controlled equipment shall be designed and constructed to handle high activity objects.
The personnel access to compartments of high dose rate or high contamination level shall be controlled by means of locking devices with interlocks and indication for actuation and unavailability.
Biological protection shall be designed in a conservative way, taking into account the build-up of radionuclides over the plant lifetime, the potential loss of shielding efficiency due to effects of interactions of neutron and gamma rays with the shielding, due to reactions with other materials, decontamination solution, and the expected temperature conditions in design basis accidents.
The choice of materials for the shield shall be made on the basis of the nature of the radiation, the shielding, mechanical and other properties of materials and space limitations.
Ventilation systems shall be installed to:
1. prevent spreading of gaseous radioactive substances in plant compartments;
2. reduce and maintain compartments' airborne concentrations below the established limits and as low as reasonably achievable in all operational states and design basis accidents;
3. cleanup the air in premises containing inert or harmful gases.
In designing a ventilation system, the following factors shall be taken into account:
1. mechanisms of thermal and mechanical mixing;
2. limited effectiveness of dilution in reducing airborne contamination;
3. exhausting of the air from areas of potential contamination at points near the source of contamination;
4. ensuring adequate distance between exhaust air discharge point and the intake point;
5. providing a higher pressure in the less contaminated zones in comparison with the zones of higher contamination level;
6. preventing the spread of fire-released smoke products to neighboring compartments.
Design shall provide for ventilation and air cleaning systems before discharge of gaseous radioactive substances to the environment.
Filters of air cleaning systems shall be sufficiently reliable to perform their function with the necessary decontamination factor in all operational modes. The design shall provide means to test their efficiency.
The requirements with regard to the classification of zones and compartments, radiation monitoring, the individual protection means and the access control are established by a different regulation.
Provisions shall be made in the design for an automated system for radiation monitoring at the workplace and at the NPP site, and a system for radiation monitoring at the radiation protection and the monitored areas. These systems shall ensure the collection and processing of information on the radiation conditions, on the effectiveness of protective barriers, on the radionuclide activity, and information necessary to predict changes in the radiation conditions in all operational states and accident conditions. Particularly design should be included measuring systems/equipment with the following capabilities:
* monitoring of radiation exposure and contamination of personnel;
* monitoring of personnel and equipment at the main exit points from controlled areas and supervised areas;
* monitoring of radioactive effluents;
* monitoring of local radiation dose rates at NPP personnel routinely accessible locations;
* indicate the radiation levels at NPP locations in accident conditions;
* measuring and analyzing of the samples taken from the plant systems or from the environment, in operational states and in accident conditions.
10. EMERGENCY PREPAREDNESS
The operating organization shall provide emergency management facilities and equipment to monitor the accident progression and manage the response. An on-site emergency control center, separated from the plant control room, shall be established.
Information about important plant parameters and radiological conditions in the plant and its immediate surroundings should be available there. The control center should provide means of communication with the control room, the supplementary control room and other important points in the plant, and with the on-site and off-site emergency response organizations.
The centre shall receive information on unit's status during the phases of accident progression and on the radiological conditions at the NPP site and its surroundings.
Appropriate measures shall be taken to protect the occupants against hazards resulting from a severe accident.
Emergency plans for protection of plant personnel and the public in case of nuclear and radiological emergencies, including severe accidents shall be developed by the operating organization in coordination with off-site authorities. The plans shall be tested periodically to demonstrate their credibility.
The operating organization shall provide emergency management facilities and equipment to monitor the accident progression and manage the response.
Instruments, tools, equipment, documentation, and communication systems for use in emergencies (including necessary mobile equipment), whether located on-site or off-site, shall be stored, maintained, tested and inspected sufficiently frequently so that they will be available and operational during DBA and DEC. Access to these storage locations shall be possible even in case of extensive infrastructure damage.
An on-site emergency control center, separated from the plant control room, shall be provided. Information about important plant parameters and radiological conditions in the plant and its immediate surroundings should be available there. The emergency control center should provide means of communication with the control room, the supplementary control room and other important points in the plant, and with the on-site and off-site emergency response organizations. Appropriate measures shall be taken to protect the occupants of the main control room, the supplementary control room and emergency control center against hazards resulting from nuclear and radiological emergencies.
11. QUALITY MANAGEMENT SYSTEM
A quality management system that describes the overall arrangements for the management, performance and assessment of the plant design shall be prepared and implemented. This programme shall be supported by more detailed plans for each structure, system and component so that the quality of the design is ensured at all times.
Design, including subsequent changes or safety improvements, shall be carried out in accordance with established procedures that call on appropriate engineering codes and standards, and shall incorporate applicable requirements and design bases. Design interfaces shall be identified and controlled.
The adequacy of design, including design tools and design inputs and outputs, shall be verified or validated by individuals or groups separate from those who originally performed the work. Verification, validation and approval shall be completed before implementation of the detailed design.
(հավելվածը խմբ. 29.09.16 թիվ 992-Ն որոշում)