Thông tư 30/2012/TT-BKHCN

CIRCULAR

THE REGULATIONS ON NUCLEAR SAFETY APPLICABLE TO THE DESIGNS OF NUCLEAR POWER PLANTS

Pursuant to the Law on Atomic Energy dated June 03, 2008;

Pursuant to the Government's Decree No. 28/2008/NĐ-CP dated March 14, 2008, defining the functions, tasks, powers and organizational structure of the Ministry of Science and Technology;

Pursuant to the Government's Decree No. 70/2010/NĐ-CP dated June 22, 2010, specifying and guiding a number of articles of the Law on Atomic Energy applicable to nuclear power plants;

At the proposal of the Director of the Vietnam Agency for radiation and nuclear safety;

The Minister of Science and Technology issues a Circular on the regulations on nuclear safety applicable to the designs of nuclear power plants

Chapter I

GENERAL REGULATIONS

Article 1. Scope of regulation

This Circular deals with the general regulations on nuclear safety applicable to nuclear power plants.

Article 2. Subjects of application

This Circular is applicable to the investors, the organizations that participate in the consultation, design, manufacture, construction, repair, maintenance, operation, design appraisal, and licensing of nuclear power plants.

Article 3. Interpretation of terms

In this Circular, the terms below are construed as follows:

1. The plant states is all possible states of the nuclear power plant, including the normal operation and abnormal operation (hereinafter referred to as operational state), the state upon the occurrence of design basis accidents and beyond design basis accidents (hereinafter referred to as accident conditions).

2. Normal operation is a state in which the nuclear power plant operates within certain operational limits and conditions. Normal operation includes start-up, power operation, reactor shutdown, maintenance, testing, and refueling.

3. Abnormal operation is a deviation from the normal operation that is expected to occur at least once during the operating lifetime of the nuclear power plant without significantly impacting the items important to safety or leading to accident conditions.

4. Design basis accidents are the accident that the design of the nuclear power plant is able to restrict upon their occurrence so that the damage to the fuel and the release of radioactive material are kept within authorized limits.

5. Beyond design basis accidents are the accidents that are more severe than the design basis accidents. The nuclear power plant might suffer damage upon the occurrence of such accidents. They are assessed to provide solutions for improving the resistance of the nuclear power plant and limit the radioactive consequences to an acceptable level.

6. Postulated initiating events are the postulated accidents directly arising out of the damage of the structure, system, parts, or operation errors, and the damage directly arising out of the internal and external hazards during the nominal power operation, low power operation, or reactor shutdown.

7. Deterministic safety analysis is the method for anticipating the phenomena likely to occur after a postulated initiating event by applying certain acceptance criteria and principles. Deterministic safety analysis including neutronic, thermal-hydraulic, radioactive, thermomechanical, and structural analysis using calculation instruments.

8. Probabilistic safety analysis is a systematic approach to determining the risks, the damage scenarios of which the probabilities are quantified by calculation instruments.

9. Severe accidents are beyond design basis accidents that cause significant damage to the reactor core.

10. Accident management is a series of actions during the evolution of a beyond design basis accidents for the purpose of:

a) Preventing the escalation of an accident into a severe accident;

b) Mitigating the consequences of a severe accident (if any);

c) Achieving a long-term safe stable state.

11. Safe state is a state of the nuclear power plant following an abnormal operation or accident in which the reactor is subcritical and the fundamental safety functions can be ensured and maintained stable for a long time.

12. Controlled state is the a of the nuclear power plant following an abnormal operation or accident in which the fundamental safety functions are still can be ensured and maintained for a time sufficient to take the measures for reaching a safe state.

13. Components are independent devices, accessories, or components of systems such as pipelines, pumps, valves.

14. A system include the parts fit together to perform a function such as the reactor system, cooling system, control system.

15. Structure is the construction designed to cover and protect, such as buildings, reactor vessels, fuel rods, or supporting structures such as shelves and suspension frame.

16. Safety system is the system that ensures the shutdown of reactors and the residual heat removal from the core, or limits the consequences of abnormal operation and design basis accidents. The safety systems include the protection system, the safety actuation system, and safety system support features such as cooling, lubrication, and power supply.

17. Safety system support features is a collection of equipment that provide services such as cooling, lubrication, and power supply for the protection system and the safety actuation system.

18. Items important to safety are the items in the safety group and of which the malfunction or failure may lead to radiation exposure of the personnel or the public.

19. Ultimate heat sink is the atmosphere, the sea, the rivers or lakes to which the residual heat of the nuclear power plant is transferred.

20. Coolant pressure boundary is the pressure parts, including:

a) The pressure vessel, the pipeline, and valves (the parts of the core cooling system);

b) The parts connected to the reactor coolant system such as the peripheral containment isolation valve at the pipe that penetrates the containment, the secondary isolation valve that is closed during the normal operation at the pipe that does not penetrate the containment, the discharge valve and safety valve of the reactor coolant system.

21. The design basis is the conditions, processes, and natural or human factors taken into account in the design of the nuclear power plant, so that the safety system of the nuclear power plant is still operational according to the design without exceeding the authorized limits upon the occurrence of such conditions, processes, and factors.

22. Safety limits are the range of the operational parameters within which the operation of the nuclear power plant is proven safe.

23. Common cause failures are failures of two or more structures, systems, and components due to a single specific event or cause.

24. Single failure is a failure that results in loss of capability of a component to perform its intended safety functions, or any consequential failure that results from the loss of capability to perform safety functions.

25. Single failure criterion is a criterion (or requirement) applied to a system in order to such that it must be capable of performing its tasks in the presence of any single failure.

26. Diversity is the presence of two or more redundant systems or components to perform a certain function. The attributes of these systems and parts are different so as to reduce the possibility of common cause failures.

27. Redundancy is the provision of alternative (identical or diverse) structures, systems and components so that any one can independently perform the same function regardless of the state of operation or failure of any other.

28. Physical separation is the separation by geometry such as distance, orientation, by appropriate barriers, or by a combination thereof.

29. ALARA principle is a radiation protection principle that ensures the radiation exposure of the personnel and the public are kept as low as reasonably achievable.

Chapter II

GENERAL REGULATIONS ON NUCLEAR SAFETY APPLICABLE TO THE NUCLEAR POWER PLANT DESIGN

Article 4. General requirements for the nuclear power plant design

1. The design of the nuclear power plant and the items important to safety must ensure that safety functions can be performed with necessary reliability. The nuclear power plant must be safely operated within the operational limits and conditions throughout its intended lifetime, can be safely decommissioned, and impacts on the environment are minimized.

2. The results of the deterministic safety analysis and the probabilistic safety analysis shall be examined to ensure that due consideration has been given to the prevention of accidents and, and to the mitigation of consequences of accidents that occur.

3. The activity and volume of the generation of radioactive waste and discharges are kept to the minimum.

4. The experience gained in the design, construction, and operation of other nuclear power plants, as well as the results of relevant research programs must be taken into account.

5. When assessing the conformity of the design with the safety requirements prescribed in this Circular, the standards established by competent State agencies, the producers’ standards, and the applicable international standards shall apply.

Article 5. Fulfillment of fundamental safety functions

1. The fundamental safety functions of the nuclear power plant include: controlling the reactivity, removing heat from the reactor and the fuel store; confining the radioactive materials, shielding against radiation, controlling planned radioactive releases, and limiting accidental releases.

2. The fundamental safety functions prescribed in Clause 1 of this Article must be fulfilled under any plant state.

3. A systematic approach shall be taken to identify:

a) The items important to safety necessary to fulfill the fundamental safety functions;

b) The inherent features that contribute to fulfilling the fundamental safety functions or affect the fundamental safety functions under any plant state.

Article 6. Radiation protection

The conditions for radiation protection in the nuclear power plant design must be satisfied, including:

1. The radiation doses to the site personnel and the public do not exceed the prescribed limits, and they are kept as low as reasonably achievable under any plant state.

2. The situations that might lead to high radiation doses or major releases of radioactive material to the environment must be practically eliminated.

3. The measure for mitigating radiological consequences of highly possible accidents must be enhanced.

Article 7. Application of defence in depth

1. The application of defence in depth is to prevent and mitigate the consequences of the accidents that could lead to harmful effects on people and the environment.

2. The levels of defence in depth shall be independent as far as is practicable. The safety of the nuclear power plant under each state must be proven when the protection level decreases.

3. Multiple physical barriers to the release of radioactive material to the environment shall be provided.

4. The failures and deviation from normal operation are minimized, and accidents are prevented as far as is practicable. The minor deviation in the parameter of the nuclear power plant does not lead to a cliff edge effect.

5. The control of plant behavior by means of inherent and engineered features must be provided such that failures and deviations from normal operation that require actuation of safety systems are minimized or excluded.

6. The safety systems must be able to be automatically activated in accident conditions.

7. The structures, systems, components, and procedures for mitigating the consequences of failures and deviations during normal operation that exceed the capability of safety systems shall be provided.

8. Multiple means for ensuring that the fundamental safety functions are performed shall be provided, thereby ensuring the effectiveness of the barriers and mitigating the consequences of any failure or deviation from normal operation

9. To ensure that the defence in depth is maintained, the following factors shall be prevented as far as is practicable:

a) The challenges to the integrity of physical barriers;

b) The failure of one or more barriers;

c) The failure of a barrier as a consequence of the failure of another barrier;

d) The harmful consequences of errors in operation and maintenance.

10. The capability of the first, or at most the second, level of defence, ensured as far as is practicable when a failure or deviation from normal operation occurs.

Article 8. Postulated initiating events

1. A systematic approach shall be adopted to identify all postulated initiating events that could result in serious consequences, and the postulated initiating events with a significant frequency of occurrence. These events must be considered in the design.

2. The postulated initiating events shall be identified based on the basis of engineering judgment, and a combination of deterministic and probabilistic assessments. The A justification of the extent of usage of deterministic safety analysis and probabilistic safety analysis shall be provided to ensure the adequacy of the list of foreseeable accidents.

3. An analysis of the postulated initiating events for the plant shall be made to establish the preventive measures and protective measures that are necessary to ensure the fulfillment of safety functions.

4. When a postulated initiating event occurs, the conditions below must be satisfied in the following order of priority:

a) The event does not produce significant effects on the safety, or only causes a change that the safety conditions are able to be automatically restored due to the inherent features of the nuclear power plant;

b) After the event, the nuclear power plant would return to the safe state by means of to the passive safety features or the capability to operate continuously of the systems that control the postulated initiating event;

c) After the event, the nuclear power plant would return to the safe state due to the actuation of safety systems;

d) After the event, the nuclear power plant would return to the safe state due to the implementation of specified processes.

5. An engineering judgment shall be provided for excluding from the design the initiating events not in the list of postulated initiating events.

6. Where prompt and reliable action would be necessary, the automatic actuation of safety systems must be designed to prevent the postulated initiating events from escalating into more severe conditions. Where prompt response would not be necessary, the safety system is activated manually, or the operator may take actions instead of activating the safety system, the following requirements must be satisfied:

a) The administrative, operational, and emergency procedures shall be adequately specified;

An assessment of the potential for an operator to worsen an event through incorrect diagnosis of the necessary recovery process or erroneous operation of equipment shall be made in order to provide suitable solutions;

c) Any equipment necessary for manual response and the recovery processes shall be placed at most suitable locations to ensure its availability at the time of need, and allow safe access to such equipment under the environmental conditions anticipated.

7. The equipment and procedures necessary for keeping control over the whole nuclear power plant and mitigating the consequences under of a loss of control shall be provided.

Article 9. Internal and external hazards

1. All foreseeable internal and external hazards, including the hazards directly or indirectly caused by human shall be identified and assessed. Hazards shall be assessed for determining the postulated initiating events and consequences in the design of items important to safety, including:

a) Internal hazards including fire, explosion, flooding, missile generation, collapse of structures; falling objects, and pipe whip;

b) Natural external hazards such as meteorological, hydrological, geological and seismic events. Human induced external hazards arising from military bases, industrial facilities, oil storage, chemical storage, transport activities.

2. In the short term, the safety of the plant shall not be dependent on off-site services such as electricity supply and fire fighting services. The period of self-protection of the nuclear power plant is determined based on the conditions off-site services conditions.

3. The seismic design of the plant must provide a sufficient safety margin to protect the nuclear power plant against seismic events and to cliff edge effects.

4. At the positions where many generating units are located, the design must take due account the potential simultaneous impacts on several generating units. The design must also take due account of the independent safety of other generating units when one of them is damaged.

Article 10. Design basis accidents

1. The conditions for identifying design basis accidents are derived from the postulated initiating events for the purpose of establishing the boundary conditions for the nuclear power plant.

2. Using the design basis accident conditions prescribed in Clause 1 of this Article to identify the design basis of the safety system and the items important to safety aiming to bringing the nuclear power plant back to the safe state and mitigating the consequences of the accidents that might occur. The conditions prescribed in Clause 1 of this Article shall be used to define the design bases for safety systems and items important to safety, with the objective of returning the plant to a safe state and mitigating the consequences of any accidents.

3. When a design basis accident occurs, the key plant parameters must not exceed the specified design limits.

4. The analysis of the design basis accidents must take due account of certain failures in safety systems, the errors of design criteria, the use of postulations, models, and input parameters.

Article 11. Beyond design basis accidents

1. The conditions for identifying beyond design basis accidents are derived on the basis of engineering judgment, deterministic assessments, and probabilistic assessments.

2. The analysis undertaken shall include identification of the features that are designed for preventing beyond design basis accidents and mitigating their consequences if they do occur. These features must:

a) Be independent, especially for more frequent accidents;

b) Capable of performing in the conditions pertaining to beyond design basis accidents and severe accidents;

c) Have reliability commensurate with the intended functions;

d) The features of containment must be able to withstand extreme scenarios, including the melting of the reactor core.

3. The possibility of significant releases of radioactive materials upon the occurrence of a beyond design basis accident must be minimized. When radioactive releases occur, it is necessary to take measures for spatially and temporally limiting the releases in order to protect the public, and sufficient time shall be made available for implementing these measures

4. If the results of the engineering judgments, deterministic safety assessments, and probabilistic safety assessments indicate that combinations of events could lead to abnormal operation or accidents, such combination shall be considered design basis accidents or part of beyond design basis accidents. Such combinations shall be considered part of the original postulated initiating events.

Article 12. Safety classification

1. The items important to safety shall be identified and classified based on their functions and importance to safety.

2. The interference between items important to safety shall be prevented, and any failure of items important to safety in a lower class must not affect the items important to safety in a higher class

3. The equipment that performs multiple functions shall be classified according to the most important function.

Article 13. Design limits

1. The design limits consistent with the key physical parameters for each item important to safety for the nuclear power plant shall be specified for all operational states and for accident conditions.

2. The design limits must be consistent with national technical regulations and standards, and requirements from competent agencies.

Article 14. Regulations on the design process

1. The organizations that design the nuclear power plant must provide for the design process of the items important to safety consistent with the relevant national standards and technical regulations, and the verified technological experience.

2. The verified designing methods and technical solutions shall be applied, ensuring that the fundamental safety functions are fulfilled for all operational state and for accident conditions.

Article 15. Design criteria for safety

1. The design of equipment must take due account of the possibility of common cause failures of items important to safety, the determination and application of the requirements of the diversity, redundancy, physical separation, and functional independence.

2. The single failure criterion shall be applied as follows:

a) The single failure criterion shall be applied to each safety group;

b) Each spurious action shall be considered to be one mode of failure with regard to a safety group or a safety system;

c) The failure of a passive component must be considered, unless it is certified to be reliable in the single failure analysis. The failure of a passive component is very unlikely and its function would remain unaffected by the postulated initiating events.

3. The fail-safe design shall be incorporated into the systems and components important to safety, so that their failure or the failure of a safety system support feature does not prevent the performance of the safety function.

Article 16. Design of items important to safety

1. The design of items important to safety must be consistent with national standards, technical regulations, and requirements from competent agencies.

2. The proven design shall be used for the items important to safety. If not, the items of high quality and of a technology that has been qualified and tested shall be used.

3. Then applying the national standards and technical regulations prescribed in Clause 1 of this Article, the applicability, conformity, and sufficiency of them must be evaluated. The application of design that ensures safety functions better than the regulations in such documents where necessary.

4. The items important to safety must be design so that they can be manufactured, constructed, and installed in accordance with the established procedure, ensure the achievement of intended features and level of safety as prescribed.

5. The design basis must specify the capability, reliability, and functionality of items important to safety for relevant operational states, for accident conditions, and for conditions arising from internal and external hazards; ensuring that the criteria are met over the lifetime of the nuclear power plant.

6. The design basis for each item important to safety shall be systematically justified and documented.

7. The items important to safety shall be designed and located to minimize the likelihood of consequences and the impacts of external hazards. Such design and location must be conformable with other safety requirements.

8. The safety system support features (including power cabling and control cabling) must be designed to minimize the impact cause by the interactions among the buildings that contain items important to safety, and other structures of the nuclear power plant in the conditions arising from external hazards.

9. The items important to safety must be capable of withstanding the impacts of external hazards considered in the design. If not other features, such as passive barriers, shall be provided to protect the nuclear power plant and to ensure that the safety functions of such items shall be performed.

10. The potential hazards caused by the interactions among the systems important to safety when they operate simultaneously shall be evaluated and prevented.

11. When analyzing the potential hazards cause by interactions among the systems important to safety, due account shall be taken of the physical interconnections and possible effects of a system on the working environment of another, to ensure that the changes in the working environment do not affect the reliability of the system.

12. 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.

13. The reliability of items important to safety shall be commensurate with their safety significance, in particular:

a) The quality of items important to safety are assessed and ensured in all stages of procurement, installation, qualification, operation, and maintenance to be capable of withstanding all conditions arising from design basis accidents:

b) In the selection of equipment, consideration shall be given to both spurious operation and possible failures. The selection of equipment of which the failures are easy to repair and replace.

14. The items important to safety must not be compromised by disturbances in the electrical power grid, including disturbances in the voltage and frequency.

Article 17. Safety systems

1. Interference between safety systems or between redundant components of a system shall be prevented by means such as physical separation, electrical isolation, functional independence and independence of data transfer.

2. The safety system equipment of the safety system, including cables and raceways, shall be readily identifiable in the plant for each redundant safety system equipment.

3. Safety systems must not be shared between generating units unless this contributes to safety enhancement.

4. Safety system support features and items related to safety may be shared between generating units, unless it would increase either the likelihood of an accident or the consequences of an accident.

Article 18. Interfaces of safety with security and safeguards

The implementation and development of safety measures, nuclear security measures, and the State system of accounting and controlling nuclear materials of the nuclear power plant shall integrated so that they do not compromise one another

Article 19. Operational limits and conditions

The operational limits and conditions shall be established in the design of the nuclear power plant, including:

1. The safety limits;

2. The limits to safety systems;

3. The operational limits and conditions for operational state;

4. The system constraints and control procedure constraints on the processes important to safety;

5. The requirements for surveillance, maintenance, testing and inspection to ensure that structures, systems and components function as intended in the design, to comply with the requirement for optimization on the ALARA principle;

6. The operational configurations, including operational restrictions in response to the failure of safety systems or systems related to safety;

7. The actions and completion times for actions in response to deviations from the operational limits and conditions.

Article 20. Calibration, testing, maintenance, repair, replacement, inspection and monitoring of items important to safety

The design of items important to safety must:

1. Facilitate their calibration, testing, maintenance, repair, replacement, inspection, and monitor of the capability of performing their functions and the maintenance of their integrity in all conditions specified in their design basis;

2. Ensure that their calibration, testing, maintenance, repair, replacement, inspection, and monitor do not cause radiation overdose to the performers;

3. Ensure that their calibration, testing, maintenance, repair, replacement, inspection, and monitor do not reduce the reliability of safety functions;

4. If the items important to safety cannot be designed to satisfy the requirements of testing, inspections or monitor to the extent desirable, a reliable technical justification must be provided that incorporates the following approaches

a) There are methods for testing, inspecting, and supervising indirectly through reference items, or using verified calculation methods; predictions shall be made for their replacements

b) There is sufficient safety margin to compensate for possible failures.

Article 21. Qualification of items important to safety

1. A qualification program for items important to safety shall be implemented to verify that items important to safety are capable of performing their intended functions in current environmental conditions and in the changes in environmental conditions considered in the design basis throughout their intended lifetime.

2. The qualification program for items important to safety must include the consideration of ageing effects caused by environmental factors, including vibration, irradiation, humidity or temperature. When the items important to safety are subject to natural external events, the qualification program for such items in similar conditions that happened before shall be considered.

3. The qualification program for items important to safety shall take due account of all unfavorable environmental conditions that could arise during the operation of the nuclear power plant.

Article 22. Ageing management

1. The design life and margins of items important to safety at a nuclear power plant shall be determined. The ageing, neutron embrittlement and degradation shall be taken into account, ensuring the capability of items to perform their necessary safety functions throughout their design life.

2. The monitor, testing, sampling, and inspection to assess ageing mechanism specified at the design stage shall be carried out. The adverse changes of the nuclear power plant or degradation that might occur during its operation shall be identified.

Article 23. Design for optimal operator performance

1. The human factors shall be systematically assessed, including human – machine interface for considering in the design.

2. The design must be consistent with the required minimum number of workers to simultaneously perform all the operations necessary to bring the plant into the safe state in abnormal operation or accidents.

3. The design must be suitable for the experience of the operators in similar nuclear power plants, and assist the operators in judging and handling the situations arising during the operation of the nuclear power plant and equipment maintenance.

4. The design must be optimal the fulfillment of the operators’ responsibilities, and limit the effects of operating errors on safety.

5. The human–machine interface design, and the information provided for operators must be sufficient and manageable, suitable for making decisions and take necessary actions.

6. The information necessary for operators includes:

a) The general state of the plant;

b) The operational limits and conditions;

c) The information about the automatic actuation of safety systems;

d) The information about the operation of systems related to safety systems;

dd) The information about the necessity and time for manual actuation of specified safety actions.

7. The working condition and environment must be designed to ensure the safety and effectiveness of operators.

8. The design must facilitate the success of operators’ even in a short period of time and under psychological impacts. The need for intervention by the operator shall be minimized, and it shall be demonstrated that the operator has sufficient time to make a decision and sufficient time to act.

9. The design must ensure that even the event affects the nuclear power plant, the environment in the main control room, the supplementary control room, and the corridor to the supplementary control room still do not threaten the safety of operators.

10. Human factors shall be assessed to confirm that necessary actions of operators are correctly performed; simulators may be used if necessary.

Article 24. Requirements of systems that contain fissile materials and radioactive materials

The systems in the nuclear power plant designed to contain fissile materials and radioactive materials shall be able to:

1. Prevent the occurrence of events that could lead to a loss of control and radioactive release to the environment;

2. Prevent criticality and overheating;

3. Keep the radioactive releases below acceptable limits and as low as reasonably achievable in any situation;

4. Mitigate the radiological consequences of accidents.

Article 25. Requirements of radioactive waste management and the decommissioning of the nuclear power plant

The requirements of radioactive waste management and the decommissioning the nuclear power plant must be consider at the design stage, including:

1. The selection of materials to minimize the generation of radioactive waste;

2. The facilities necessary for the treatment and storage of radioactive waste generated during the operation and decommissioning of the nuclear power plant;

3. The easy access and the means for handling.

Article 26. Safety system support features

1. Support service systems that ensure the operability of equipment being part of a system important to safety shall be classified.

2. The reliability, redundancy, diversity, and independence of safety system support features must be commensurate with the significance to safety of the system being supported.

3. The failure of safety system support features must not simultaneously affect the redundant components of the safety system or a system that fulfill safety functions, and affect the capability to fulfill safety functions of these systems.

Article 27. Requirements of escape routes

1. The nuclear power plant must be provided with a sufficient number of escape routes, clearly marked, with emergency lighting, ventilation and other conditions essential to the safe use of these escape routes in an emergency.

2. The escape routes from the nuclear power plant must satisfy the requirements of competent State management agencies for radiation zones, fire protection, industrial safety, and nuclear power plant security.

3. At least one escape route shall be available from a workplace or an occupied area in an accident or even simultaneous accidents.

Article 28. Requirements of communication system

1. The diverse means of communication must be provided for internal and external communication. Those means must be appropriately located and available in any circumstances.

2. Suitable alarm systems shall be provided to give warnings and instructions in abnormal operation and accident conditions.

Article 29. Requirements of access to the nuclear power plant and prevention of unauthorized acts

1. The nuclear power plant shall be isolated from its surroundings with a suitable zoning. Various structural systems shall be provided so that access to the plant can be controlled.

2. The zoning prescribed in Clause 1 of this Article must take due account of access to the nuclear power plant in accident conditions for taking emergency response measures.

3. The zoning prescribed in Clause 1 of this Article must take due account of the control of access to the plant and prevention of unauthorized access to or interference with the equipment of the plant, especially the items important to safety.

Chapter III

REQUIREMENTS FOR THE DESIGN OF SPECIFIC SYSTEMS

Section 1. Reactor core and associated features

Article 30. Requirements of fuel elements and fuel assemblies

1. Fuel elements and fuel assemblies must ensure the period of use.

2. When assessing the quality of fuel elements and fuel assemblies after a period of use, the following factors must be taken into account:

a) The expansion and deformation;

b) The external pressure of the coolant;

c) The internal pressure of fission products and helium accumulation;

d) The effects of irradiation;

dd) The variation in the pressure and temperature resulting from the changes in the power of the nuclear power plant;

e) The chemical effects;

g) The static and dynamic loading; the flow-induced vibrations and mechanical vibrations;

g) The variations in the performance of heat transfer resulting from distortions or chemical effects.

3. The limits on the permissible leakage of fission products from the fuel shall be established so that the fuel remains usable below such limits.

4. Fuel elements and fuel assemblies must be able to withstand the effects related to the installation, dismantlement, transportation, and storage as prescribed.

Article 31. Requirements of the cooling capability and control rods

The geometrical design of fuel elements, fuel assemblies, and supporting structures must ensure that the capability of cooling is maintained, and the insertion of the control rods into the reactor core is not impeded in operational states and in accident conditions, other than severe accidents.

Article 32. Control of neutrons in the reactor core

1. The neutron flux distribution in the reactor core must be inherently stable in all operational state, including the states arising after shutdown, and during or after refueling, and states arising from abnormal operation; the quality of the reactor core does not degraded.

The need for the control system for maintaining the shapes, levels and stability of the neutron flux within specified design limits in all operational states shall be minimized.

2. Means of monitoring the neutron flux distributions in the reactor core shall be provided for the purpose of ensuring that the neutron flux in the core does not exceed the design limits.

3. The design of reactivity control devices must take into account the degradation of equipment caused by the effects of irradiation, burnup, changes in physical properties and production of gas.

4. The maximum degree of positive reactivity and rate of reactivity increase in operational states and accident conditions shall be limited or compensated.

5. to maintain the capability for cooling and to prevent any significant damage to the reactor core. prevent any resultant failure of the pressure boundary of the reactor coolant systems, to maintain the capability for cooling and to prevent any significant damage to the reactor core. The quality reactor core must be assured over the lifetime of the nuclear power plant. Failure of coolant pressure boundary must be prevented. The capability of cooling must be maintained, and significant damage to the reactor core must be prevented.

Article 33. Reactor shutdown

1. Means shall be provided to ensure the capability to shut down the reactor in any situation, even when the positive reactivity of the reactor is maximum.

2. The effectiveness, speed of action and shutdown margin must ensure that that the design limits for fuel are not exceeded.

3. When assessing the effectiveness of the means of shutdown of the reactor, consideration shall be given the failures arising in the plant that could make part of the means of shutdown inoperative or that could result in a common cause failure.

4. The means for shutting down the reactor must satisfy the following requirements:

a) There are at least two diverse and independent systems for the purpose of excluding common cause failures. At least on of the two reactors shutdown systems must be capable of keeping the reactor subcritical by an adequate margin and with high reliability;

b) The increase in reactivity leading to unintentional criticality during refueling, shutdown, or while the reactor is in the shutdown state must be prevented.

5. Instrumentation shall be provided and tests shall be regularly carried out for ensuring that the means of shutdown are always ready in any plant state.

Section 2. REACTOR COOLENT SYSTEM

Article 34. Requirements for the reactor coolant system

1. The components of the reactor coolant systems must satisfy the requirements of design and manufacture quality, material quality, and inspection during their operation.

2. The pipework connected to the pressure boundary must be designed to limit the leak of coolant through interfacing systems, and to prevent the release of radioactive coolant.

3. The flaws must be prevented from initiating, and any flaws that are initiated must be promptly detected, and the extension of flaws must be prevented.

4. The embrittement of the components of the coolant pressure boundary must be avoided.

5. A failure of a component contained inside the coolant pressure boundary, such as pump impellers and valves, must not cause damage to other components that are important to safety, in all operational states and in accident conditions, with due account taken of their degradation.

Article 35. Overpressure protection of the coolant pressure boundary

The operation of the pressure relief devices must be ensured protect the coolant pressure boundary against overpressure at any position, and prevent the release of radioactive materials directly to the environment.

Article 36. Control of reactor coolant

1. The volume, temperature, and pressure of the reactor coolant must not exceed the design limits in any operational state of the nuclear power plant, with due account taken of the volumetric change and leakage.

2. Adequate facilities shall be provided for the removal of activated corrosion products and fission products deriving from the fuel.

3. The capability of the facilities prescribed in Clause 2 of this Article shall be based on the design limits on permissible leakage of the fuel, with a sufficient margin to ensure a low level of circuit activity, and that the radiation releases are below the acceptable limits and are as low as reasonably achievable.

Article 37. Removal of residual heat from the reactor core

Means shall be provided for the removal of residual heat from the reactor core in the shutdown state of the reactor such that the design limits for fuel, the coolant pressure boundary and structures important to safety are not exceeded.

Article 38. Emergency cooling of the reactor core

1. Means of cooling the reactor core shall be provided. The cooling of the fuel under accident conditions shall be restored and maintained even if the integrity of the coolant pressure boundary is not maintained.

2. The means of cooling prescribed in Clause 1 of this Article must satisfy the following requirements:

a) The limiting parameters for the integrity of the fuel cladding are not exceeded;

b) The chemical reactions are kept to an acceptable level;

c) The means of cooling of the reactor core are effective and able to compensate for changes in the fuel and in the internal geometry of the reactor core;

d) The cooling of the reactor core will be ensured for a sufficient time.

3. The leak detection systems, interconnection and isolation systems, suitable redundancy and diversity shall be provided to fulfill the requirements in Clause 2 of this Article with high reliability for each postulated initiating event.

Article 39. Heat transfer to an ultimate heat sink

Systems shall be provided to transfer residual heat from items important to safety to an ultimate heat sink with high reliability for all plant states.

Section 3. CONTAINMENT

Article 40. Features of the containment system

Features of the containment system:

1. Confinement of radioactive material and radiation shielding in any situation;

2. Protection of the reactor against natural and human-induced external hazards.

Article 41. Control of radioactive releases from the containment

1. The containment shall be designed to ensure that the radioactive release from the nuclear power plant to the environment is kept as low as reasonably achievable, and below the acceptable limit.

2. The containment structure, the systems and components affecting the leaktightness of the containment system shall be designed and constructed so that the leak rate can be tested at the containment design pressure during the operation of the plant.

3. The design of the penetrations through the containment must satisfy the following requirements:

a) The number of penetrations through the containment must be kept to a minimum. The features and other requirements applicable the penetrations shall be kept at a level similar to that of the containment itself.

b) The penetrations shall be able to with stand the impacts caused by pipe movement and pipe whip, or accidents relevant to missiles, external or internal events.

Article 42. Isolation of the containment

1. Each line that penetrates the containment as part of the coolant pressure boundary or that is connected directly to the containment atmosphere shall:

a) Be automatically and reliably sealable in the event of an accident;

b) The leaktightness prescribed in Point a of this Clause shall be fitted with at least two containment isolation valves or check valves arranged in series (usually, a valve is located inside and the other is located outside the containment and shall be provided with suitable leak detection systems. The isolation valves and check valves shall be located as close to the containment as possible; each valve must be capable of independent and reliable actuation, and be periodically tested;

c) The fulfillment of the requirements prescribed in Points a and b of this Clause may not necessary for the instrument lines, or in cases in which the implementation of such requirements reduces the reliability of the safety system that includes the penetrations through the containment.

2. Each line that penetrates the containment and is neither part of the coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve located outside the containment and as close to the containment as possible.

Article 43. Access the containment

1. The access to the containment for operators shall be through airlocks. The airlocks shall be equipped with doors that are interlocked to ensure that at least one of the door is closed in any situation. The entry and the corridors must be supervised. Provisions for ensuring the safety of personnel shall be specified in the design.

2. Containment openings for the movement of equipment or material shall be designed to be closed quickly and reliably in the event that isolation of the containment is required.

Article 44. Control of containment conditions

1. Systems shall be provided to control the pressure, temperature, and buildup of fission products, or other gaseous, liquid, and solid substances that might be released inside the containment and that could affect the operation of systems important to safety.

2. The flow routes between separate compartments inside the containment must be sufficient, and designed such that the pressure differentials occurring during pressure equalization in accident conditions do not result in unacceptable damage to the pressure bearing structure or systems important to the mitigation of accident consequences.

3. The capability to remove heat from the containment must be ensured to reduce and maintain pressure and temperature inside the containment as low as possible after an accident that release high-energy fluids. The systems that remove heat from the containment shall have high reliability and redundancy.

4. The systems that control fission products shall be provided to reduce the release of fission products to the environment when in accident conditions.

5. The systems that control the concentration of hydrogen, oxygen, and other substances in the containment shall be provided to prevent fire and explosion in accident conditions.

6. The use of coverings, thermal isolations, and coatings for components and structures inside the containment must to ensure the fulfillment of the safety functions even if they are damaged.

Section 4. MEASUREMENT AND CONTROL SYSTEM

Article 45. Measurement systems

1. The measurement systems must be capable of determining the values of key parameters that could affect the fission process, the integrity of the reactor core, the reactor coolant system and the containment in order to ensure the reliable operation and safety of the nuclear power plant, determining the nuclear power plant in accident conditions, and making decisions for the purposes of accident management.

2. The measurement equipment must provide sufficient information to monitor the plant state and the course of accidents, to predict the locations of release and the amount of radioactive material that could be released and for post-accident analysis.

Article 46. Control systems

The control systems must be reliable and suitable for limiting the relevant process variables within the specified operational ranges.

Article 47. Protection system

1. The protection system must:

a) Be capable of detecting the unsafe conditions and activating safety systems automatically to achieve and maintain the safety conditions for the nuclear power plant;

b) Be capable of overriding the unsafe actions of the control systems;

c) Be capable of recovering the safety conditions of the plant even in the event of failure of the protection system itself;

d) Be capable of activating the safety system, maintaining the automated actions within an appropriate period of time after the occurrence of abnormal operation or accident conditions before the intervention by operators;

dd) Provide information for operators for monitoring the effects of automatic actions.

2. The design must be able to prevent the operators from compromising the effectiveness of the protection system, but must not impede the correct actions of operators in accident conditions.

Section 48. Reliability and testability of measurement and control systems

1. The measurement and control systems for items important to safety shall be designed for high reliability and periodic testability commensurate with their safety functions.

2. The measurement and control systems must be designed to facilitate the test, automatically notify system failures, and automatically justify failures; must be provided with functional diversity and diversity in concepts of operation to ensure the fulfillment of safety functions in any situation.

3. The safety system must be designed to permit periodic testing of their functionality when the plant is in operation, including the possibility of testing channels independently for the detection of failures and sufficient redundancy; to permit The design shall permit testing of the functions of the sensor, the input signal, the actuation mechanism, and the display.

4. When a safety system, or part of a safety system, has to be taken out of service for testing, clear directives on this suspension for testing must be provided.

Article 49. Use of computer-based equipment in systems important to safety

1. If a system important to safety is dependent on computer-based equipment, appropriate provisions for the development and testing of computer hardware and software shall be established and implemented throughout the lifetime of the system, especially the relation to software development. The entire development shall be subject to a quality management system.

2. The computer-based equipment in systems important to safety and safety related systems must:

a) Use software and hardware of high quality commensurate with the importance of the system to safety;

b) Have the entire design process, including the control, testing, and commissioning of design changes systematically documented. These documents must be regularly reviewed;

c) Be assessed by the experts who are independent from the design team and providers to ensure the high reliability;

d) Apply the principle of diverse design to equipment important to safety of which the high reliability is not clearly justified;

dd) Take due account of common cause failures derived from computer software;

e) Be protected against damage upon the when the system operation is interfere or in accident conditions.

Article 50. Separation of protection systems and control systems

1. The protection system and control system must be functionally independent, and the interference between which shall be prevented by means of separation.

2. If signals are shared by both the protection system and the control system, the signals must be classified as part of the protection system. The separation of these two systems in this case must be clearly justified.

Article 51. Control room

1. The control room must be designed to be safely operated in all operational state either automatically or manually, and measures must be taken to maintain the safe state of the nuclear power plant or to bring the nuclear power plant back to the safe state after abnormal operation and accident conditions.

2. The control room and the external environment must be appropriately separated. Adequate information must be provided to protect the personnel in the control room from the hazards resulting from accident conditions such as high radiation levels, release of radioactive materials, fire, explosion, or toxic gases.

3. The accidents inside and outside the control room that could affect the continuous operation of the control room must be analyze, and appropriate measures shall be taken to mitigate the consequences of the accidents that do occur.

Article 52. Supplementary control room.

1. A supplementary control room with measurement and control equipment that is physically, electrically, and functionally separate from the control room prescribed in Article 51 of this Circular shall be kept available.

2. The supplementary control room must be capable of maintaining the shut down state of the reactor, removing residual heat, and monitoring the changes of the parameters when the ability to perform the related features in the control room is loss.

3. The requirements of the protection of personnel prescribed in Clause 2 Article 51 also apply to the supplementary control room.

Article 53. Emergency control center

1. The nuclear power plant must have an on-site emergency control center that is separate from the control room and the supplementary control room.

2. The information about the important the plant parameters, the radiological conditions at the nuclear power plant, and the immediate surroundings must be displayed in the center.

3. The center must be provided with means of communication with the control room, the supplementary control room, and other important locations in the plant, and with emergency response units.

4. Appropriate measures shall be taken to protect the personnel of the emergency control centre for a protracted time against hazards resulting from accident conditions.

5. The emergency control centre must be provided with necessary systems and services to permit protract the working time of emergency response personnel.

Section 5. Emergency power supply

Article 54. Emergency power supply

1. The emergency power supply at the nuclear power plant must be capable of supplying the necessary power in abnormal operation and accident conditions that cause the loss of off-site power.

2. The design basis for the emergency power supply at the nuclear power plant shall take due account of the postulated events and the associated safety functions to determine the requirements for capability, availability, required duration of the power supply, capacity and continuity.

3. The combination of emergency power supply sources such as water, steam or gas turbines, diesel engines or batteries, must be reliable and conformable with the requirements for power supply of safety systems; the design must facilitate the testing of the system functions.

4. The design of diesel engines and generators that provide emergency power supply for items important to safety must satisfy the following requirements:

a) The capability of fuel oil storage and supply systems must satisfy the demand within a specified period of time;

b) The generators are able to start and function under any condition and at any time;

c) Auxiliary systems of the generators such as cooling systems must be provided.

Section 6. SUPPORTING SYSTEMS AND AUXILIARY SYSTEMS

Article 55. Performance of supporting systems and auxiliary systems

The design of supporting systems and auxiliary systems must ensure that the performance of these systems is consistent with the importance to safety of the systems or components that they serve.

Article 56. Heat transport systems

The systems and components of the nuclear power plant that operate all the time (even in accident conditions) must be provided with auxiliary systems to remove heat. The auxiliary parts of the heat transport system must be isolated.

Article 57. Process sampling systems and post-accident sampling systems

1. Process sampling systems and post-accident sampling systems shall be provided for promptly determining the concentration of radionuclides in fluid process systems, and in gas and liquid samples taken from systems or from the environment, in all operational states and in accident conditions.

2. Appropriate measures shall be taken for monitoring the activity in gas fluid systems that have the potential for radioactive contamination, and for the sampling.

Article 58. Compressed air systems

In the design basis, the quality, flow rate, and cleanness of the air supplied for the compressed air systems shall be specified.

Article 59. Air conditioning systems and ventilation systems

1. Appropriate air conditioning, air heating and ventilation systems shall be provided in the nuclear power plant to maintain the environmental conditions necessary for the systems and components important to safety.

2. The capability of air cleaning of ventilation systems in buildings must satisfy the following requirements:

a) The radioactive release within the nuclear power plant is below the prescribed limits;

b) The concentration of airborne radioactive substances in the areas accessed by personnel is below the prescribed limits;

c) The level of airborne radioactive substances within the nuclear power plant is below the limits and kept as low aw reasonably achievable;

d) The ventilation in the rooms that contain inert gases or noxious gases does not impair the capability to control radiation;

dd) The release of gaseous radioactive material to the environment is below acceptable limits and kept as low as reasonably achievable.

2. Lower pressure (partial vacuum) shall be maintained in areas where radioactive contamination is higher, and in accessible areas.

Article 60. Fire protection systems

1. Fire protection systems, including fire detection and fire extinguishing systems, fire containment barriers, and smoke control system shall be provided throughout the nuclear power plant, with due account taken of the fire hazard analysis results.

2. The fire protection system at the nuclear power plant must be capable of dealing with all fire scenarios.

3. The fire extinguishing systems must be capable of automatic actuation where appropriate. The design and location of fire extinguishing systems must ensure that their abnormal operation would not significantly impair the capability of items important to safety.

4. The fire detection systems must provide information for operators about the location and spread of the fire when it occurs.

5. The fire detection systems and fire extinguishing system in postulated initiating events must be able to resist the effects of these events.

6. Non-combustible and heat resistant materials shall be used wherever practicable throughout the plant, especially in the containment and the control room.

Article 61. Lighting systems

Adequate lighting shall be provided in all operational areas of the nuclear power plant in operational states and in accident conditions.

Article 62. Overhead lifting equipment

1. The items important to safety and other items in the proximity the items important to safety shall be lifted and lowered by equipment.

2. The design overhead lifting equipment shall:

a) Prevent the lifting of excessive loads;

b) Prevent unintentional dropping;

c) Safely move itself and the items being lifted and lowered;

d) Have safety interlocks;

dd) Provided with anti-earthquake design if they are used in the areas where items important to safety are located.

Section 7. POWER CONVERSION SYSTEMS

Article 63. Steam supply system, water supply system and generators

1. The design of the steam supply system, water supply system and turbine generators shall ensure that the coolant pressure boundary is not exceeded in any situation.

2. The steam supply system must have qualified steam isolation valves capable of closing in any situation.

3. The design and capacity steam supply system and water supply system shall be able to prevent abnormal operation from escalating to accident conditions.

4. Turbine generators must be designed to resist vibration and overspeed. Appropriate measures shall be taken for minimize the effects turbine-generated missiles on items important to safety

Section 8. SYSTEMS FOR TREATMENT OF RADIOACTIVE EFFLUENTS AND RADIOACTIVE WASTE

Article 64. Systems for treatment and control of waste

1. Systems shall be provided for treating solid radioactive waste and liquid radioactive waste at the nuclear power plant to keep the amounts and concentrations of radioactive releases below the authorized limits and as low as reasonably achievable.

2. Systems and facilities shall be provided for the management and storage of radioactive waste in the nuclear power plant for a period of time consistent with the disposal option.

3. The design of the plant shall incorporate appropriate features to facilitate the transport and handling of radioactive waste. Consideration shall be given to the access of equipment that lift and pack radioactive waste.

Article 65. Systems for treatment and control of liquid and gaseous effluents

1. Systems shall be provided for treating and residual liquid and gaseous radioactive effluents to keep their amounts below the authorized limits, and as low as reasonably achievable.

2. Liquid and gaseous radioactive effluents shall be treated at the plant so that exposure of the public due to discharges to the environment is as low as reasonably achievable.

3. The design of the plant shall incorporate suitable means to keep the release of liquid radioactive effluents to the environment below the authorized limits and as low as reasonably achievable.

4. The cleanup equipment for the gaseous radioactive substances shall provide the necessary retention factor to keep radioactive releases below the authorized limits. The efficiency for filter systems must be testable. The performance and function of these systems must be regularly monitored over their lifetime. The filter cartridges must be able to be replaced while the air is passing.

Section 9. FUEL HANDLING AND STORAGE SYSTEMS

Article 66. Fuel handling and storage systems

1. The fuel handling and storage systems shall be provided at the nuclear power plant to maintain the fuel control at all times during fuel handling and storage.

2. The design of the nuclear power plant must facilitate the lifting, movement and handling of fresh fuel and spent fuel.

3. The design of the nuclear power plant must prevent significant damage to the items important to safety during the transfer of fuel or casks, or in the event of fuel or casks being dropped.

4. The systems of handling and storing fresh fuel and spent fuel must:

a) Prevent crititicality by physical means or by means of physical processes or by a specified margin, and preferably by use of appropriate geometrical configurations;

b) Facilitate the inspection of fuel;

c) Facilitate the maintenance, periodic inspection of components important to safety;

d) Prevent damage to the fuel;

dd) Prevent the dropping of fuel in transit;

e) Provide identification for individual fuel assemblies;

g) Provide proper means for radiation protection;

h) Provide adequate operating procedures and a system of accounting for, and control of nuclear fuel.

5. The systems of handling and storing spent fuel must:

a) Allow removal of heat from the fuel in any situation;

b) Prevent causing unacceptable handling stresses on fuel elements or fuel assemblies;

c) Prevent the dropping of fuel in transit;

d) Prevent the potential dropping of heavy objects that damage fuel;

dd) Keep suspect or damaged fuel elements or fuel assemblies safe;

e) Control levels of soluble absorber if this is used for criticality safety;

g) Facilitate the maintenance and disassembly of fuel handling and storage facilities;

h) Facilitate decontamination of fuel handling and storage areas and equipment;

i) Accommodate all fuel take from the reactor core in accordance with the plan for core management;

k) Facilitate the transit of fuel form the storage and the preparation for off-site transport.

6. When the nuclear power plant uses water pools for fuel storage, the design of the plant must be able to:

a) Control the temperature, chemical properties, and activity of the water use for handling or storing spent fuel;

b) Monitor and control the water level in the pools, and detect leakage;

c) Prevent the exposure of fuel elements and fuel assemblies in the pool due to pipe break.

Section 10. RADIATION PROTECTION

Article 67. Design for radiation protection

1. Provision shall be made for ensuring that radiation doses to personnel at the nuclear power plant will be maintained below the dose limits and will be kept as low as reasonably achievable.

2. Radiation sources in the plant and radiation risks associated with them shall be identified. The exposure form these sources shall be kept as low as reasonably achievable. The integrity of the fuel cladding shall be maintained. The generation, development, and effects of corrosion products and activation products shall be controlled

3. Materials used in the manufacture of structures, systems and components shall be selected to minimize activation of the material.

4. Technical measures shall be taken to prevent the release of radioactive substances, radioactive waste, and radioactive contamination in the nuclear power plant.

5. The design of the nuclear power plant must ensure that the access of personnel to areas with potential exposure and areas of possible radioactive contamination is controlled. Exposures and radioactive contamination shall be prevented or minimized by controlling means and ventilation systems.

6. The nuclear power plant shall be divided into zones according to the level of radiation and radioactive contamination in the operation of the nuclear power plant (including refueling, maintenance and inspection); the areas of potential radiation and contamination in accident conditions shall be identified. Shielding shall be provided so that radiation exposure is prevented or minimized.

7. The design of the nuclear power plant must ensure that the doses received by the personnel during the during normal operation, refueling, maintenance and inspection are kept as low as reasonably achievable, special equipment shall be used to meet these requirements.

8. The equipment subject to frequent maintenance or manual operation shall be located in areas of low dose rate to minimize the exposure of workers.

9. Facilities shall be provided for the decontamination of operating personnel and plant equipment.

Article 68. Means of radiation monitoring

1. Equipment shall be provided to monitor radiation in operational states and design basis accidents. If possible, equipment shall be provided to monitor radiation in beyond design basis accidents.

2. Stationary dose rate meters shall be provided for monitoring local radiation dose rates at plant locations that are routinely accessible by personnel, and where the changes in radiation levels in operational states allow the access of personnel for certain specified periods of time.

3. The stationary dose rate meter shall:

a) Display the radiation levels at the necessary plant locations in accident conditions;

b) Provide sufficient information n the control room and control positions so that personnel can intervene if necessary.

4. The stationary radiation monitors shall be capable of measuring the activity of radioactive substances in the atmosphere in areas routinely occupied by personnel and where the radioactivity might be such as to necessitate protective measures. The detected concentration of radionuclides shall be displayed in the control room and other necessary positions. Radiation monitors shall be installed areas subject to possible contamination as a result of equipment failure or other unusual circumstances.

5. Stationary equipment and laboratories shall be provided for promptly determining the radiation concentration in fluid process systems, in gas and liquid samples taken from plant systems or from the environment in operational states and in accident conditions.

6. Stationary equipment shall be provided for monitoring radioactive effluents and effluents with possible contamination during discharges from the plant to the environment

7. Instruments shall be provided for measuring surface contamination. Stationary monitors, including portal radiation monitors, hand and foot monitors, shall be provided at the exits from controlled areas and supervised areas to control the radiation for operating personnel and equipment

8. Facilities shall be provided for monitoring exposure and radioactive contamination of operating personnel, for the purposes of assessing and recording the cumulative doses to personnel over the period of working at the nuclear power plant.

9. The exposure and other radiological impacts in the vicinity of the plant shall be assessed by observing the dose rates or environmental radioactivity, with particular reference to:

a) Exposure pathways to people, including the food chain;

b) Radiological impacts, if any, on the local environment;

c) The possible accumulation or radioactive substances in the environment;

d) The possibility of unauthorized radioactive releases.

Chapter IV

REGULATIONS ON THE IMPLEMENTATION

Article 69. Effects

1. This Circular takes effect after 45 days from the date on which it is signed.

2. Organizations and individuals are recommended to send feedbacks on the difficulties arising during the course of implementation to the Ministry of Science and Technology for amendment and supplementation./.