Circular No. 30/2012/TT-BKHCN stipulates nuclear safety requirements for the design of nuclear power plants.

CIRCULAR

Regulations on nuclear safety requirementsfor nuclear power plant design

Pursuant to the Atomic Energy Law dated June 3, 2008;

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

Pursuant to Decree No. 70/2010/NĐ-CP dated June 22, 2010 of the Government detailing and guiding some articles of the Atomic Energy Law regarding nuclear power plants;

On the proposal of the Director of the Radiation and Nuclear Safety Agency;

The Minister of Science and Technology issues this Circular stipulating nuclear safety requirements for nuclear power plant design.

PART I
GENERAL PROVISIONS

Article 1. Scope of Regulation

This Circular stipulates general nuclear safety requirements for nuclear power plant design (hereinafter referred to as NPP).

Article 2. Applicability

This Circular applies to investors and agencies and organizations participating in the advisory, design, manufacturing, construction, repair, maintenance, operation, design review, and construction permit issuance processes for NPPs.

Article 3. Explanation of Terms

In this Circular, the following terms are understood as follows:

1. Status of NPP is a general term referring to all possible statuses of NPP including normal operating status and abnormal status (commonly referred to as operational status), design accident status and beyond-design accident status (commonly referred to as accidents).

2. Normal operation is the status in which the NPP operates within defined operating limits and conditions. Normal operation includes startup, power operation, reactor shutdown, maintenance, inspection, and fuel reloading.

3. Abnormal status (referred to as abnormality) is an event deviating from normal operation that is predicted to occur at least once during the entire operating period of the NPP but does not significantly affect important safety items and does not cause an accident.

4. Design accident is an accident considered as a condition for design ensuring that the NPP can withstand such accidents when they occur, so that fuel damage and radioactive material release remain below the limits set by the competent authority.

5. Beyond-design accident is a more serious accident than a design accident. The NPP may be damaged when such an accident occurs, and it is evaluated to enhance the NPP's resilience capability and limit radiation consequences to permissible levels.

6. Postulated initiating accident is an assumed accident directly resulting from structural, system, component failure, or operational error and directly caused by internal and external hazards when the NPP operates at rated power, low power, or reactor shutdown state.

7. Deterministic safety analysis is a method predicting phenomena occurring after a postulated initiating accident through the application of specific rules and acceptance criteria. Deterministic safety analysis includes neutron, thermal-hydraulic, radiation, thermo-mechanical, and structural analyses using computational tools.

8. Probabilistic safety analysis is a systematic and comprehensive approach to determine risks and failure scenarios with quantified occurrence probabilities using computational tools.

9. Serious incident are beyond-design accident causing significant core damage.

10. Accident management is a series of actions implemented throughout the course of a beyond-design accident aimed at the following objectives:

a) Preventing the development of the accident into a severe one;

b) Minimizing the consequences of a severe accident if it occurs;

c) Achieving a long-term stable safe state.

11. Safe state is the state of the NPP after abnormalities or accidents occur, where major safety functions are maintained and stabilized over a long period with the reactor in sub-limit state.

12. Controlled state is the state of the NPP after abnormalities or accidents occur, where major safety functions are maintained and stabilized for a sufficient time to implement measures to achieve a safe state.

13. Component may be an independent device or a part of a system such as a pipe, pump, or valve.

14. System" consists of components assembled together to perform a function such as the reactor system, cooling system, control system.

15. Structure is a building structure serving to shield and protect such as buildings, reactor vessel, spent fuel pool, or supporting structures like supports and frames.

16. Safety system is a system ensuring reactor shutdown, heat removal from the core, or limiting the consequences of abnormal status and design accidents. Safety systems include protection systems, safety feature actuation systems, and support systems such as cooling, lubrication, and power supply.

17. Safety system support system is a system of equipment supporting cooling, lubrication, and power supply for protection systems and safety feature actuation systems.

18. Important safety item is an item belonging to the safety category or an item whose malfunction or failure could lead to radiation exposure to staff and the public.

19. Final heat sink environment is the air, sea, river, or lake environment serving to remove excess heat from the NPP.

20. Pressurized component boundary includes pressurized components comprising:

a) Pressure vessel, pipes, pumps, and valves (components of the reactor core cooling system);

b) Components connected to the reactor cooling system such as outer containment isolation valves at pipe penetrations through the containment, second isolation valves typically closed during normal operation at non-penetrating pipes, relief and safety valves of the reactor cooling system.

21. Design entity includes conditions, processes, factors caused by natural or human activities taken into account in designing the NPP, so that when these conditions, processes, and factors occur, the NPP's safety systems still operate according to design, and safety limits approved by the competent authority are ensured.

22. Safety limit is the range of operating parameters where the operation of the NPP has been proven to be safe.

23. Common cause failure is the failure of two or more structures, systems, and components caused by the same accident or cause.

24. Single failure Failure is when a system or part loses its ability to perform its designed safety function, or when subsequent failure originates from the loss of the ability to perform that safety function.

25. Single failure criterion is the criterion (or requirement) applied to a system to ensure that the system can still perform its function even with a single failure.

26. Diversity is the presence of two or more systems or parts to perform the same defined function. These systems or parts have different attributes to minimize the likelihood of common cause failures.

27. Reserve is the provision of structures, systems, and components (similar or different) capable of substituting each other to independently perform the same function without depending on the operational status or failure of one structure, system, or component among them.

28. Physical separation is geometric separation such as distance, orientation, or separation by appropriate barriers, or a combination of both methods.

29. ALARA principle is the principle ensuring radiation protection so that the dose received by radiation workers and the public is kept as low as reasonably achievable.

Chapter II
GENERAL REQUIREMENTS FOR NUCLEAR SAFETY FOR

NUCLEAR POWER PLANT DESIGN

Article 4. General requirements for nuclear power plant design

1. The design of the nuclear power plant and important safety items must ensure the ability to perform safety functions with the required reliability. The nuclear power plant can operate safely within the operating limits and conditions throughout its design lifetime, be dismantled safely, and minimize environmental impact.

2. Consider the results of deterministic safety analysis and probabilistic safety analysis to ensure that the design has considered preventing accidents and mitigating accident consequences if they occur.

3. Ensure that the activity level, radioactive waste volume generated, and radioactive emissions are at the minimum level.

4. Take into account the experience gained during the design, construction, and operation of other nuclear power plants, as well as the results of relevant research programs.

5. When evaluating the compliance of the design with the safety requirements stipulated in this circular, if argumentation is necessary, apply the standards issued by the competent state authority and the manufacturer's standards, international standards permitted for application in Vietnam.

Article 5. Ensuring primary safety functions

1. The primary safety functions of the nuclear power plant include: reactivity control; heat removal from the reactor and fuel storage areas; containment of radioactive materials, shielding of radiation, control of radioactive emissions according to design, and limitation of radioactive dispersion incidents.

2. Ensure all primary safety functions specified in Clause 1 of this Article for all states of the nuclear power plant.

3. Use a systematic approach to determine:

a) The essential safety items needed to ensure primary safety functions;

b) The inherent characteristics contributing to the assurance of primary safety functions or affecting primary safety functions in all states of the nuclear power plant.

Article 6. Radiation Protection

Ensure radiation protection conditions when designing the nuclear power plant, including:

1. The dose received by workers at the nuclear power plant and the public does not exceed the limit set by the competent authority and ensure the ALARA principle in all states of the nuclear power plant.

2. Prevent situations that could lead to the release of high-dose or large quantities of radioactive material into the environment.

3. Enhance measures to mitigate the consequences of radioactive releases from highly probable incidents.

Article 7. Multi-layer protection requirements

1. Implement multi-layer protection requirements to prevent and mitigate the consequences of incidents that may harm people and the environment.

2. Protection layers must always be maintained and must be as independent as possible. When reducing the level of protection, it must be demonstrated that safety for the nuclear power plant is still ensured for each specific state.

3. There are multiple physical protection layers to prevent the release of radioactive material into the environment.

4. Minimize the possibility of failures and deviations during normal operation, preventing incidents from occurring to the maximum extent possible. Small deviations in nuclear power plant parameters shall not lead to sudden escalation effects.

5. Control systems for the nuclear power plant must have technical advantages and inherent characteristics such that they can minimize or eliminate the need to activate safety systems due to failures or deviations during normal operation.

6. Safety systems must have the capability to automatically start in the event of an incident.

7. Structures, systems, components, and procedures must be established to minimize the consequences arising from failures or deviations during normal operation that the safety system cannot control.

8. Multiple means must be provided to perform key safety functions, ensure the effectiveness of protection layers, and minimize the consequences of failures or deviations during normal operation.

9. Maintain multi-layer protection requirements by preventing to the maximum extent the following factors:

a) Impact on the integrity of physical protection layers;

b) Failures of one or more protection layers;

c) Failure of one protection layer due to failure of another layer;

d) Consequences of operational and maintenance errors.

10. Ensure to the maximum extent the protective capability of the first layer or, if there is damage to protection layers, up to the second layer, when there are failures or deviations during normal operation.

Article 8. Hypothetical Incidents

1. Apply a systematic approach to identify all hypothetical incidents that could result in serious consequences and those with high frequency. These incidents must be considered in the design.

2. Hypothetical incidents are determined based on technical arguments, combining deterministic and probabilistic assessments. The scope of deterministic and probabilistic safety analysis is justified to ensure the comprehensiveness of the list of predictable incidents.

3. Establish necessary preventive and protective measures for performing safety functions through the analysis of hypothetical incidents.

4. In the event of a hypothetical incident, the following conditions in priority order must be met:

a) The incident does not significantly impact safety or only causes changes that can be self-recovered due to the inherent characteristics of the nuclear power plant;

b) After the incident, the nuclear power plant must return to a safe state due to passive safety features or the continuous operation of hypothetical incident control systems;

c) After the incident, the nuclear power plant must return to a safe state due to the activation of safety systems;

d) After the incident, the nuclear power plant must return to a safe state due to the application of predefined procedures.

5. Technical justification must be provided to exclude hypothetical incidents not included in the list of hypothetical incidents at the design stage.

6. For cases requiring rapid and reliable response, automatic activation of safety systems must be designed to prevent hypothetical incidents that could lead to more severe situations. For cases where a rapid response is not required, the activation of safety systems is performed by humans or operators can perform actions instead of activating safety systems, then the following requirements must be followed:

a) Properly define administrative, operational, and emergency response procedures;

b) Evaluate the potential for equipment failure, incorrect operations, or erroneous judgments by operators during the necessary recovery process that could worsen the situation to develop appropriate solutions;

c) Equipment necessary for manual emergency response and recovery processes must be placed in suitable locations to ensure readiness, timeliness, and safe access to this equipment under anticipated environmental conditions.

7. Necessary equipment and procedures must be provided to maintain overall control of the nuclear power plant and minimize consequences in case of loss of control.

Article 9. Internal and External Hazards

1. Identify and assess all internal and external hazards that could occur at the nuclear power plant (NMĐHN), including those caused directly or indirectly by human actions. When designing critical safety components, evaluate hazards to determine assumed initiating events and consequences, including:

a) Internal hazards include fire, explosion, flooding, projectile objects, structural collapse, falling objects, pipeline impact;

b) External hazards caused by natural factors such as meteorology, hydrology, geology, earthquakes. External hazards caused by human factors such as military facilities, industrial sites, oil storage, chemical storage, transportation activities.

2. The NMĐHN must have the ability to ensure short-term safety independently without relying on external support (electricity, firefighting). The duration of self-sustained safety for the NMĐHN is determined based on external support conditions.

3. Seismic design must have sufficient safety margins to protect against earthquake hazards and sudden rise and fall effects on the NMĐHN.

4. At locations with multiple units, design must consider the possibility of simultaneous impact hazards affecting multiple units. Design must also consider the independent safety capability of other units when one unit is affected by a hazard.

Article 10. Design Incidents

1. Determine design incident conditions from assumed initiating events to establish boundary conditions for the NMĐHN.

2. Use the design incident conditions specified in Clause 1 of this Article to establish design bases for safety systems and critical safety components with the aim of returning the NMĐHN to a safe state and minimizing consequences when incidents occur.

3. In the event of a design incident, key parameters of the NMĐHN must not exceed specific design limits.

4. Design incident analysis must take into account the possibility of certain failures in safety systems, design criterion errors, and the use of assumptions, models, and input parameters.

Article 11. Beyond-Design-Basis Incidents

1. Determine beyond-design-basis incident conditions based on technical justification, deterministic safety assessment, and probabilistic safety assessment.

2. Analyze to identify design characteristics aimed at preventing beyond-design-basis incidents or mitigating their consequences when they occur. These characteristics must possess the following attributes and capabilities:

a) Independence, particularly when applied to frequently occurring incidents;

b) Capability to maintain operation under beyond-design-basis incident conditions, severe incidents;

c) Reliability commensurate with designed function;

d) For reactor containment, design characteristics must ensure resistance to extreme scenarios, including core melt in the reactor.

3. Limit the likelihood of large-scale radioactive material release during beyond-design-basis incidents. In the event of radioactive release, measures must be taken to limit the release area and time to protect the public and provide sufficient time to implement these measures.

4. If technical justification, deterministic safety assessment, and probabilistic safety assessment results indicate that a combination of events could lead to operational anomalies or incidents, then consider this combination as a design incident or part of a beyond-design-basis incident. Consider this combination as part of the initial assumed initiating event.

Article 12. Safety Classification

1. Identify and classify safety critical items based on their functions and levels of safety importance.

2. Prevent cross-influence between safety critical items so that failures of lower-level safety critical items do not affect higher-level safety critical items.

3. Equipment with multiple functions shall be classified according to its most important function.

Article 13. Design Limits

1. Determine appropriate design limits for each safety critical item based on key physical parameters for operational states and during incidents.

2. Design limits must comply with technical standards, national standards, and regulations of competent authorities.

Article 14. Design Process Regulations

1. Design organizations of nuclear power plants must establish design process regulations for safety critical items in accordance with relevant technical standards, national standards, and verified technological experience.

2. Apply verified design methods and technical solutions to ensure that primary safety functions are maintained for operational states and during incidents.

Article 15. Safety Criteria in Design

1. The design of equipment must consider the possibility of common cause failures among safety critical items, determine and apply requirements for diversity, redundancy, physical separation, and functional independence.

2. Apply single failure criteria as follows:

a) Single failure criteria shall be applied to each safety group.

b) Each action outside permitted procedures shall be considered as a potential failure type for a safety group or safety system.

c) Failures of passive components must also be considered unless they are assessed as highly reliable when applying single failure analysis. Passive component failures must be unlikely and their function should not be affected by assumed initiating events.

3. The principle of inherent safety design shall be applied to safety systems and components to ensure that safety functions are performed even if these systems or supporting safety systems fail.

Article 16. Design of Safety Critical Items

1. Design safety critical items in compliance with technical standards, national standards, and regulations of competent authorities.

2. Apply verified designs for safety critical items. If this cannot be met, high-quality items with evaluated and tested technology must be used.

3. When using technical standards, national standards, and regulations stipulated in Clause 1 of this Article, it is necessary to determine and evaluate these documents regarding applicability, suitability, and completeness. Higher quality designs ensuring higher safety functions than those prescribed in these documents may be applied if deemed necessary.

4. Design safety critical items so that they can be manufactured, constructed, and installed in accordance with established procedures, ensuring compliance with design characteristics and safety levels as prescribed.

5. Design bases must specify the capabilities, reliability, and functions of safety critical items for related operational states, during incidents, and from internal and external hazards; ensuring compliance with prescribed criteria throughout the lifecycle of the nuclear power plant.

6. Design bases for each safety critical item must be systematically documented and substantiated.

7. Safety critical items must be designed and installed in positions minimizing the likelihood and impact of external hazards. Such design and installation must still comply with other safety requirements.

8. Supporting safety systems (including electrical cables and control cables) must be designed to minimize the impact of interactions between buildings containing safety critical items and other structures of the nuclear power plant during external hazards.

9. It must be ensured that safety critical items can withstand the effects of external hazards considered in the design. Otherwise, additional features such as passive protective layers must be provided to protect the nuclear power plant and ensure the performance of safety functions of these items.

10. Potential hazards arising from interactions between safety critical systems when operating simultaneously must be assessed and prevented.

11. When analyzing potential hazards arising from interactions between safety critical systems, physical connections and possible impacts of one system on the working environment of another system must be considered to ensure that environmental changes do not affect the reliability of the system.

12. In cases where two systems containing safety critical liquids are connected and operate at different pressures, both systems must be designed to withstand higher pressure or there must be provisions to prevent exceeding the design pressure of the system operating at lower pressure.

13. Reliability of safety critical items must be ensured at a level commensurate with their safety importance, demonstrated through the following requirements:

a) Safety critical items must be evaluated and assured in terms of quality throughout procurement, installation, acceptance, operation, and maintenance stages to ensure their ability to withstand incidents as designed.

b) When selecting equipment, unintended actions and potential failures must be considered. Priority should be given to equipment with easily repairable or replaceable failure modes.

14. Safety critical items must not be affected by disturbances from the power grid, including voltage and frequency disturbances.

Article 17. Safety System

1. Measures must be taken to prevent the interaction effects between safety systems or between backup devices or components of the same system through physical separation, electrical isolation, functional independence, and data transmission independence.

2. Equipment of the safety system, including cables and wiring conduits within the nuclear power plant for each backup device of the safety system, must be marked for easy identification.

3. Units shall not share safety systems unless this contributes to increasing the level of safety.

4. Support systems for safety systems and safety-related items may be shared among units, except where such sharing increases the likelihood or consequences of an accident.

Article 18. Compatibility of Safety with Security and Inspection

Measures for the construction and implementation of nuclear safety, security, and management systems for auditing and controlling nuclear materials in nuclear power plants must be integrated so that they do not adversely affect each other.

Article 19. Operational Limits and Conditions for Safety

Operational limits and conditions for safety must be established during the design of the nuclear power plant, including the following requirements:

1. Establishing safety limits;

2. Establishing limits for safety systems;

3. Establishing operational limits and conditions for operating states;

4. Establishing system limits and control process limits for all important safety processes;

5. Requiring monitoring, maintenance, testing, and inspection to ensure that structures, systems, and components perform their designed functions, meet optimization requirements, and comply with the ALARA principle;

6. Establishing operational configurations, including operational limits in case of failure of safety systems or safety-related systems;

7. Determining actions and time limits for taking corrective actions when deviations from operational limits or conditions occur.

Article 20. Calibration, Testing, Maintenance, Repair, Replacement, Inspection, and Monitoring of Important Safety Items

The design of important safety items must ensure the following requirements:

1. Facilitating calibration, testing, maintenance, repair, replacement, inspection, and monitoring of their ability to perform functions and maintain integrity under all conditions defined in the design basis;

2. Ensuring that calibration, testing, maintenance, repair, replacement, inspection, and monitoring activities do not expose personnel to excessive radiation doses;

3. Ensuring that calibration, testing, maintenance, repair, replacement, inspection, and monitoring activities do not reduce the reliability of safety functions;

4. Where it is not possible to design important safety items to allow direct testing, inspection, or monitoring at the desired level, reliable technical justification must be provided using the following approaches:

a) Indirect testing, inspection, or monitoring methods through reference items, using verified calculation methods capable of prediction, can replace those items;

b) Adequate safety margins must be available to compensate for potential failures.

Article 21. Ensuring the quality of important safety components

1. Implement a quality assessment program for important safety components to confirm that these components have the ability to perform their necessary functions under current environmental conditions and anticipated changes in environmental conditions throughout their design life cycle.

2. The quality assessment program for important safety components must include consideration of aging effects caused by environmental factors, including vibration, radiation exposure, humidity, and high temperatures. When important safety components are affected by external hazards of natural origin, the quality assessment program for those components must be considered under similar conditions that have occurred.

3. The quality assessment program for important safety components must take into account all adverse environmental conditions that may arise during the operation of nuclear power plants.

Article 22. Aging Management

1. Determine the design life and reserve capacity of important safety components taking into account aging, neutron irradiation embrittlement, and quality degradation; ensure that these components can perform their required safety functions throughout their designed operational life.

2. Conduct monitoring, testing, sampling, and inspection to evaluate the aging mechanisms identified at the design stage, as well as to identify any adverse changes or quality degradation occurring during the operation of the nuclear power plant.

Article 23. Optimal Design for Operator Actions

1. Systematically assess human factors including human-machine interaction to incorporate these factors into the design.

2. The design must be compatible with the minimum number of operators required to simultaneously perform actions necessary to bring the nuclear power plant to a safe state in case of abnormalities or incidents.

3. The design must be compatible with the experience of operators at similar nuclear power plants, supporting operators in recognizing and handling situations during the operation of nuclear power plants and equipment maintenance.

4. The design must optimize the performance of operators' responsibilities, minimizing the impact on safety due to operator errors.

5. For human-machine interface design, information provided to operators must be comprehensive and easy to manage, suitable for decision-making and performing necessary actions.

6. Information needed for operators includes:

a) General status of the plant;

b) Operating limits and conditions;

c) Information on the automatic activation of safety systems;

d) Information on the operation of systems related to safety systems;

e) Information on the necessity and timing for manual activation of predetermined safety actions.

7. Working conditions and environments must be designed to ensure safety and efficiency for operators.

8. The design must facilitate successful operator actions even under time-limited and psychologically stressful conditions; minimize the need for operator actions; if operator actions are necessary, it must be demonstrated that the time available is sufficient for decision-making and action execution.

9. The design must ensure that, even if an incident affects the nuclear power plant, it does not cause the main control room, auxiliary control room, and corridors leading to the auxiliary control room to become unsafe for operators.

10. Assess human characteristics to confirm that necessary operator actions are performed accurately; use simulation equipment in the assessment (if necessary).

Article 24. Requirements for Systems Storing Fissile Materials and Radioactive Substances

The systems within the Nuclear Power Plant (NMĐHN) designed to store fissile materials or radioactive substances must have the following features:

1. Prevent the possibility of incidents that could lead to loss of control and release of radioactive material into the environment;

2. Prevent the possibility of reaching critical states and overheating;

3. Maintain radioactive release below prescribed limits in all situations and adhere to the ALARA principle;

4. Minimize radiological consequences when incidents occur.

Article 25. Requirements for Radioactive Waste Management and Decommissioning of NMĐHN

Requirements for radioactive waste management and decommissioning of NMĐHN must be considered from the design stage, including the following contents:

1. Selection of materials to minimize the generation of radioactive waste;

2. There must be facilities necessary for processing and storing radioactive waste generated during operation and decommissioning of NMĐHN;

3. Access must be easy and there must be necessary processing means available.

Article 26. Support Systems for Safety Systems

1. Classify service support systems ensuring the operational capability of equipment as part of important safety systems.

2. Support systems for safety systems must have reliability, redundancy, diversity, and independence commensurate with the level of importance of the safety systems they support.

3. Malfunctions of support systems for safety systems shall not simultaneously affect the backup parts of safety systems or safety function execution systems and impact their ability to perform safety functions.

Article 27. Requirements for Emergency Egress

1. NMĐHN must have sufficient emergency egress routes with clear signage, lighting, ventilation, and other essential conditions to be usable in emergencies.

2. Emergency egress routes from NMĐHN must meet the requirements of the competent state management agency regarding radiation zones, fire and explosion protection, industrial safety, and nuclear power plant security.

3. Workplaces and areas where people are present must have at least one usable emergency egress route in case of incidents, even if multiple incidents occur simultaneously.

Article 28. Requirements for Communication Systems

1. There must be diverse communication means capable of internal and external communication. These means must be located appropriately and usable in all situations.

2. There must be appropriate alarm systems to warn and guide in abnormal situations and during incidents.

Article 29. Requirements for Entry and Exit of NMĐHN and Prevention of Unauthorized Acts

1. Isolate NMĐHN from surrounding areas according to suitable planning, with different structural systems to control entry and exit of the plant.

2. The planning specified in Clause 1 of this Article must consider access to NMĐHN in case of incidents and implementation of emergency response measures.

3. The planning specified in Clause 1 of this Article must consider the ability to control movement and prevent unauthorized access or interference with plant equipment, especially those of significant safety importance.

Chapter III
REQUIREMENTS FOR SPECIFIC SYSTEMS DESIGN

Section 1
REACTOR OPERATING ZONE AND RELATED CHARACTERISTICS

Article 30. Requirements for Fuel Rods and Fuel Assemblies

1. Fuel rods and fuel assemblies must ensure integrity in all situations, capable of withstanding radiation and conditions within the reactor core region, even when their quality deteriorates after a period of use.

2. The factors to be considered when assessing the quality of fuel rods and fuel assemblies after a period of use include:

a) Expansion and deformation;

b) External pressure from coolant;

c) Internal pressure caused by fission products and helium accumulation;

d) Radiation effects;

đ) Changes in pressure and temperature due to changes in nuclear power plant capacity;

e) Chemical effects;

g) Static and dynamic loads; vibrations due to flow and mechanical oscillations;

h) Changes in thermal conductivity due to deformation or chemical effects.

3. Establish limits on the leakage of fission products from the fuel such that below these limits, the fuel may continue to be used.

4. Fuel rods and fuel assemblies must be able to withstand impacts related to installation, dismantling, transportation, and storage as prescribed.

Article 31. Requirements for Cooling Capacity and Operation of Control Rods

The geometric design of fuel rods, fuel assemblies, and supporting structures must ensure maintenance of cooling capacity and not impede the insertion of control rods into the reactor core during normal operation as well as in case of incidents, except in severe accidents.

Article 32. Neutron Control in the Reactor Core Region

1. The neutron flux distribution in the active region must be inherently stable in all operating states including post-shutdown, during or after fuel loading, under abnormal conditions, and during incidents; without degrading the quality of the reactor core region.

Maximize the limitation on the necessity to use the control system to maintain the shape, level, and stability of the neutron flux within the designed limits in all operating states.

2. There must be means to monitor the neutron flux distribution in the reactor core region, ensuring that the neutron flux in the active region does not exceed the designed limit.

3. The design of the reactivity control devices must take into account the degradation of equipment quality due to radiation effects, combustion processes, physical property changes, and generated gases.

4. Limit or compensate for maximum positive reactivity and reactivity increase rate during operation and in case of incidents.

5. Ensure the quality of the reactor core region throughout the life cycle of the nuclear power plant, prevent boundary damage to coolant pressure vessels, maintain cooling capacity, and prevent significant damage to the reactor core region.

Article 33. Shutdown of the Reactor

1. Means must be provided to ensure the ability to shut down the reactor in all situations, including when the reactor has the highest positive reactivity.

2. The effectiveness, speed, and shutdown reserve must be ensured to keep the fuel design limits from being exceeded.

3. When evaluating the effectiveness of reactor shutdown means, consider all failures in the nuclear power plant that could partially disable the shutdown means or lead to common cause failures.

4. Reactor shutdown means must meet the following requirements:

a) At least two independent systems with different characteristics to eliminate the possibility of common cause failures. At least one of the two shutdown systems must have the capability to maintain sub-limit status with high reserves and reliability;

b) Prevent an increase in reactivity leading to an undesired critical state during fuel loading, reactor shutdown, or when the reactor is in a shutdown state.

5. Special equipment and regular inspections must be provided to ensure that reactor shutdown means are always ready at any state of the nuclear power plant.

Section 2
REACTOR COOLING SYSTEM

Article 34. Requirements for the Reactor Cooling System

1. Components of the reactor cooling system must meet requirements regarding design quality, manufacturing quality; material quality, and inspection requirements during operation.

2. The pressure boundary cooling fluid piping must be designed and manufactured to prevent cooling fluid leakage at connections, avoiding the release of radioactive cooling fluid.

3. Cracks shall not occur, and cracks, when they appear, must be promptly detected; cracks shall not develop further once they occur.

4. The components of the pressure boundary cooling fluid shall not become brittle due to material embrittlement.

5. Damage to internal components of the pressure boundary cooling fluid, such as pump impeller damage, valve damage, shall not lead to the destruction of other safety-critical components in all operating states and accidents, taking into account the degradation of their quality.

Article 35. Overpressure Protection for the Pressure Boundary Cooling Fluid

Ensure the operation of depressurization equipment to protect against overpressure at all locations of the pressure boundary cooling fluid, preventing direct radioactive release from the nuclear power plant into the environment.

Article 36. Control of the Reactor Coolant

1. Control the quantity, temperature, and pressure of the reactor coolant to ensure that it does not exceed design limits in all operating states of the nuclear power plant, considering changes in volume and coolant leakage.

2. There must be a system to remove corrosion products and fission products released from the fuel.

3. The capability of the system specified in Clause 2 of this Article must be based on design limits for fuel leakage rates with sufficient margin to ensure low primary circuit radioactivity levels; ensuring radioactive emissions below permitted limits and adhering to the ALARA principle.

Article 37. Removal of Excess Heat from the Active Zone of the Reactor

There must be a method to remove excess heat from the active zone of the reactor in shutdown conditions to ensure fuel, pressure boundary cooling fluid, and important safety structure design limits.

Article 38. Emergency Cooling of the Active Zone of the Reactor

1. There must be a method to cool the active zone, restore and maintain fuel cooling even if the integrity of the pressure boundary cooling fluid cannot be maintained during an accident.

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

a) Parameters related to fuel cladding integrity shall not be exceeded;

b) Chemical reactions shall be kept at acceptable levels;

c) The active zone cooling method must be effective, capable of addressing changes in fuel and internal geometry within the active zone;

d) Active zone cooling must be ensured for the required duration.

3. There must be a leak detection system, connection parts, isolation, and appropriate redundancy systems to reliably implement the requirements set out in Clause 2 of this Article for each assumed initiating event.

Article 39. Heat Transfer to the Final Heat Sink

There must be a reliable system to transfer excess heat from important safety components to the final heat sink in all states of the nuclear power plant.

Section 3
REACTOR VESSEL

Article 40. Features of the Reactor Containment System

The reactor containment system design shall have the following features:

1. Retaining radioactive materials and shielding radiation in all situations;

2. Protecting the reactor from external hazards caused by natural or human factors.

Article 41. Control of Radioactive Release from the Reactor Containment

1. The reactor containment design must comply with the ALARA principle for radioactive release from the nuclear power plant to the environment, and this level must be lower than the permissible limit.

2. The design, manufacture, and installation of structures, systems, and components affecting the containment system's sealability shall ensure that leakage testing at design pressure can be conducted throughout the operation of the nuclear power plant.

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

a) The number of penetrations through the reactor containment must be kept to the minimum. Other features and requirements for penetrations must be at the same level as the reactor containment design;

b) Penetrations must have the ability to withstand forces caused by pipe movement, impact, or other incidents related to radioactive material, internal and external hazards.

Article 42. Isolation of the Reactor Containment

1. For pipes penetrating the reactor containment that are part of the pressure boundary of the coolant or directly connected to the air inside the reactor containment, the following requirements must be met:

a) Must have high reliability automatic closure capability when an incident occurs;

b) The closure capability specified in Point a of this Clause is ensured through at least two containment isolation valves or check valves installed in series (usually one inside and one outside the reactor containment), along with a leak detection system. The isolation valves or check valves must be placed as close to the reactor containment as possible; each valve must have independent, reliable start-up capability and be regularly tested;

c) It may not be necessary to implement the requirements set forth in Points a and b of this Clause for pipes containing measuring equipment or in cases where applying those provisions would reduce the reliability of the safety system with penetrations through the reactor containment.

2. For pipes penetrating the reactor containment that are not part of the pressure boundary of the coolant or are not directly connected to the air inside the reactor containment, only one isolation valve placed outside the reactor containment as close as possible is required.

Article 43. Entry and Exit of the Reactor Containment

1. Personnel access doors to the reactor containment must be airtight. These doors operate on an interlocking mechanism to ensure that at least one door remains closed in all situations. There must be a monitoring system designed for the doors and passageways. Design requirements related to ensuring personnel safety must be included.

2. The opening of the reactor containment for equipment or material transport must be designed so that it can be quickly and reliably closed when containment isolation is needed.

Article 44. Control of Conditions within the Reactor Containment

1. There must be a system for controlling pressure, temperature, and monitoring the accumulation of fission products or other gases, liquids, or solids released in the reactor containment that could affect the operation of important safety systems.

2. There must be sufficient connections between separate compartments within the reactor containment, designed such that pressure differences arising from pressure equalization during an incident do not result in unacceptable damage to pressure-bearing structures or important systems for mitigating the effects of the incident.

3. Ensure the ability to remove heat from the reactor containment to reduce and maintain the internal pressure and temperature of the reactor containment at an acceptable low level after an incident releasing high energy flow. The system performing this heat removal function must be highly reliable and redundant.

4. There must be a system for controlling fission products aimed at reducing the amount of fission products released into the environment in the event of an incident.

5. There must be a system for controlling hydrogen, oxygen, and other concentrations in the reactor containment during an incident to prevent explosions.

6. The use of coatings, insulation layers, and platings for components and structures inside the reactor containment must ensure that safety functions are not affected, even if these layers are damaged.

Section 4
MEASUREMENT AND CONTROL SYSTEM

Article 45. Measurement System

1. The measurement system must be capable of measuring the values of key parameters that may affect the fission process, the integrity of the reactor core, the reactor cooling system, and the containment to ensure reliable and safe operation of the nuclear power plant (NPP), determine the status of the NPP in case of an incident, and make decisions for incident management purposes.

2. Recording devices must provide sufficient information to monitor the status of the NPP and the progression of incidents, predict the source and amount of radiation released, and for post-incident analysis.

Article 46. Control System

The control system must have sufficient reliability and suitability to limit related process variables within the defined operating range.

Article 47. Protection System

1. The protection system must meet the following requirements:

a) Possess the ability to detect unsafe conditions and automatically activate safety systems to achieve and maintain safe conditions for the NPP;

b) Possess superior capability to address unsafe operations on the control system;

c) Possess the ability to restore safe conditions of the NPP even when the protection system itself is damaged;

d) Possess the ability to trigger the operation of safety systems, maintain automatic operations for a reasonable period after abnormalities occur or during incidents, before operator intervention is required;

đ) Provide information to operators to enable them to monitor the impact of automatic actions.

2. Design must prevent operator actions from harming the effectiveness of the protection system, but not hinder correct operator actions during incidents.

Article 48. Reliability and Testability of Measurement and Control Systems

1. The design of measurement and control systems for important safety items must have high reliability and corresponding periodic testability commensurate with the safety function of those items.

2. Measurement and control systems must be designed to facilitate testing, automatically report system errors, and automatically correct errors; they must have functional diversity and operational principles to maintain safety functions in all situations.

3. Safety systems must be designed to allow periodic testing of system functions, including when the NPP is in operation, including the ability to independently test channels to detect malfunctions and to maintain redundancy; permit functional testing of sensor devices, input signals, actuator mechanisms, and display screens.

4. When stopping the operation of safety systems or parts of safety systems for testing, clear indications of such testing stops must be provided.

Article 49. Use of Computer-Based Equipment in Important Safety Systems

1. If important safety systems depend on computer-based equipment, there must be provisions for continuous improvement and testing of hardware and software throughout the system's lifecycle, particularly concerning software upgrade processes. A quality management system must cover the entire improvement process.

2. Computer-based equipment in safety systems and systems related to safety must meet the following requirements:

a) Use high-quality hardware and software corresponding to the level of importance of the system for safety;

b) Systematically document the entire design process, including control, inspection, and trial operation when design changes occur. This documentation must be regularly reviewed;

c) Independently evaluated by experts separate from the design team and supplier to ensure high reliability;

d) Apply diversity design principles for safety-critical equipment where high reliability cannot be clearly demonstrated;

đ) Consider common cause failures originating from computer software;

e) Be protected against damage due to interference during system operation or incidents.

Article 50. Separation of Protection Systems and Control Systems

1. The protection system and control system must be designed to be functionally independent, preventing mutual influence through separation measures.

2. If the protection system and control system share signals, these signals must be grouped as part of the protection system. The separation of the two systems in this case must be clearly justified.

Article 51. Control Room

1. The design of the control room must ensure that it can operate safely in automatic or manual mode during all operational states of the nuclear power plant (NMĐHN), and must have measures to maintain the NMĐHN in a safe state or return it to a safe state following abnormalities and incidents.

2. There must be barriers and appropriate separation measures between the control room and the external environment. Sufficient information must be provided to protect control room staff from hazards such as high radiation levels, radioactive material dispersion, explosions, or toxic gas spread when incidents occur.

3. Analyze internal and external incidents that may affect the continuous operation of the control room and implement practical measures to mitigate the consequences of incidents when they occur.

Article 52. Auxiliary Control Room

1. An auxiliary control room with measuring and control equipment physically, electrically, and functionally separated from the main control room specified in Article 51 of this Circular must be provided.

2. The auxiliary control room must have the capability to maintain a safe shutdown state, thermal load excess, and monitor parameter changes when the main control room loses its related functions.

3. Apply the employee protection requirements stipulated in Clause 2 of Article 51 to the auxiliary control room.

Article 53. Emergency Operation Center

1. The nuclear power plant must have an emergency operation center at a separate location from the main control room and the auxiliary control room.

2. Information on important parameters in the nuclear power plant, radiation conditions at the nuclear power plant, and the surrounding environment must be displayed at the center.

3. Equip the center with communication means with the main control room, auxiliary control room, other critical positions within the nuclear power plant, and emergency response units.

4. Implement measures to protect employees working at the center for extended periods from dangers during incidents.

5. Provide systems, equipment, and necessary conditions at the center to allow extended work hours for emergency responders.

Section 5
EMERGENCY POWER SUPPLY SYSTEM

Article 54. Emergency Power Supply System

1. The emergency power supply system at the nuclear power plant must have sufficient capacity to provide necessary electricity when there are abnormalities and incidents causing loss of external power supply to the nuclear power plant.

2. The design basis for the emergency power supply system at the nuclear power plant must consider hypothetical incidents and related safety functions to determine requirements for capacity, availability, required power supply time, power rating, and continuity.

3. The combination of emergency power sources such as steam turbines, diesel generators, or batteries must have reliability and selection suitable for the power supply requirements of the safety system; must be designed conveniently for functional testing of the system.

4. Design for diesel engines and emergency power generators for critical safety items must meet the following requirements:

a) Fuel storage capacity and supply systems must meet usage needs for a specific period of time;

b) The generator's ability to start and operate under all circumstances and at any time;

c) There must be auxiliary systems for the generator such as cooling systems.

Chapter 6
SUPPORT SYSTEMS AND AUXILIARY SYSTEMS

Article 55. Capacity of Support Systems and Auxiliary Systems

The design of support systems and auxiliary systems must ensure that their capacity is appropriate to the safety importance of the system or part they support or assist.

Article 56. Heat Load Systems

Nuclear Power Plant (NPP) systems and components that operate continuously (including during incidents) must have auxiliary heat load systems. Parts of the heat load system must be isolated.

Article 57. Process Sampling System and Post-Incident Sampling System

1. There must be a process sampling system and post-incident sampling system to promptly determine the concentration of radioactive material in liquid processing systems, in gas and liquid samples taken from systems or environments, in all operating conditions and during incidents.

2. Measures must be in place to monitor activity levels in liquid and gas systems that may become contaminated; sampling measures must be implemented.

Article 58. Compressed Air System

In the design basis, the quality, flow rate, and purity of the air supplied to the compressed air system must be determined.

Article 59. Air Conditioning and Ventilation Systems

1. There must be air conditioning, heating, cooling, and ventilation systems suitable for areas within the NPP to maintain necessary environmental conditions for systems and critical safety components.

2. The ventilation system in buildings must have the ability to filter air to meet the following requirements:

a) Radioactive dispersion within the NPP below the prescribed limit;

b) Radioactivity concentration in the air in areas where operational staff need to enter and exit must be below the prescribed limit;

c) Radioactivity level in the air within the NPP below the prescribed limit and comply with the ALARA principle;

d) Ventilation for rooms containing inert gases or toxic gases does not affect the ability to control radioactivity;

e) Control of radioactive gas dispersion into the environment below the prescribed limit and comply with the ALARA principle.

2. Maintain low pressure (local vacuum) in high radiation areas compared to lower radiation areas, in areas where staff enter and exit.

Article 60. Fire Protection System

1. A fire protection system, including detection and firefighting systems, fire isolation barriers, and smoke exhaust systems, must be present at all locations within the NPP, taking into account fire hazard analysis results.

2. The fire protection system at the NPP must be capable of responding to all fire scenarios.

3. The firefighting system must have the capability to automatically start at necessary locations. The design and location of the firefighting system must ensure that when there is an abnormality in this system, it does not significantly impact critical safety items.

4. The fire detection system must provide rapid information to operational staff about the location and scale of the fire as soon as it starts.

5. The fire detection system and firefighting system designed to prevent fires in simulated incident scenarios must have sufficient capability to withstand impacts from these incidents.

6. Non-combustible and insulating materials must be used at all possible locations within the NPP, particularly in the reactor containment and control room.

Article 61. Lighting System

All work areas in the nuclear power plant must be illuminated at all operational states and during emergencies.

Article 62. Hoisting Equipment

1. Important safety components and other components near the positions of important safety components must be hoisted using equipment.

2. Hoisting equipment must be designed with the following features:

a) Prevent overloading during hoisting;

b) Prevent dropping accidents;

c) Have the ability to safely move itself and the hoisted components;

d) Have interlocking safety locks;

đ) Be designed to withstand shocks if they are used in areas where important safety components are located.

Section 7
ENERGY CONVERSION SYSTEM

Article 63. Steam Supply, Water Supply, and Turbine Generator Systems

1. The design of the steam supply, water supply, and turbine generator systems must ensure that the design limit of the cooling medium pressure boundary is not exceeded under any circumstances.

2. The steam supply system must have isolation valves that have been quality tested and have appropriate closing speeds, capable of closing under any situation as prescribed.

3. The steam supply and water supply systems must be designed with suitable capacity and the ability to prevent minor operational abnormalities from developing into incidents.

4. The turbine generator must be designed to protect against vibration and overspeed. At the same time, measures must be taken to prevent the impact of objects ejected from the turbine on important safety components.

Section 8
WASTE STREAM AND RADIOACTIVE WASTE TREATMENT SYSTEM

Article 64. Waste Treatment and Control System

1. There must be a radioactive solid and liquid waste treatment system at the nuclear power plant to keep the amount and concentration of radioactive emissions below the specified limits and adhere to the ALARA principle.

2. The design of the system, management facilities, and storage of radioactive waste at the nuclear power plant must be suitable for the disposal plan within an appropriate timeframe.

3. The design of the nuclear power plant must have characteristics that facilitate the transportation and treatment of radioactive waste; consideration must be given to the ease of entry and exit, lifting, and packaging of radioactive waste.

Article 65. Liquid and Gas Waste Treatment and Control System

1. There must be systems for treating radioactive liquid and gas waste to keep residual radioactive waste below the specified limits and adhere to the ALARA principle.

2. Radioactive liquid and gas waste must be treated immediately at the nuclear power plant so that the radiation dose to the public caused by this type of waste after release into the environment adheres to the ALARA principle.

3. The design of the nuclear power plant must include appropriate measures to ensure that radioactive liquid waste released into the environment remains below the specified limits and adheres to the ALARA principle.

4. Radiation gas filtration equipment must have the necessary deposition coefficient to keep radioactive dispersion below the specified level. The system's filter performance must be testable. The operation and function of this system must be monitored regularly throughout the filter's lifecycle. The filter box must be replaceable while maintaining air flow.

Section 9
FUEL HANDLING AND STORAGE SYSTEM

Article 66. Fuel Handling and Storage System

1. There must be a fuel handling and storage system at the nuclear power plant to maintain fuel control throughout the handling and storage period.

2. The design of the nuclear power plant must have characteristics that facilitate the lifting, movement, and handling of unused and spent fuel.

3. The design of the nuclear power plant must prevent significant impacts on important safety components during movement or in the event of fuel or container drops.

4. The handling and storage system for unused and spent fuel must meet the following requirements:

a) Prevent criticality through physical means, physical processes, or safe reserves, preferably by using appropriate geometric configurations;

b) Facilitate fuel inspection;

c) Facilitate regular maintenance and inspection of important safety components;

d) Prevent fuel damage;

đ) Prevent fuel drops during transport;

e) Have identification markings for each fuel bundle;

g) Have radiation protection measures;

h) Have appropriate operating procedures and nuclear fuel auditing and control systems.

5. The handling and storage system for spent fuel must meet the following requirements:

a) Allow heat removal from the fuel in all situations;

b) Prevent excessive stress on fuel rods or bundles beyond the specified limits;

c) Prevent the possibility of fuel drops during transport;

d) Prevent the possibility of heavy objects falling and damaging the fuel;

đ) Safely store damaged or suspected damaged fuel rods or bundles;

e) Control the amount of neutron absorber dissolved in water if it is used to maintain a subcritical state;

g) Facilitate maintenance and dismantling of fuel handling and storage equipment;

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

i) Have sufficient capacity for all fuel removed from the reactor core according to planned reactor core management;

k) Facilitate the transport of fuel out of storage and preparation before transporting fuel out of the plant.

6. For nuclear power plants that use water pools for fuel storage, the design must include the following technical solutions:

a) Control temperature, chemical properties, and radioactivity of the water for processing or storing spent fuel;

b) Monitor and control water levels in the pool and leak detection measures;

c) Prevent the exposure of fuel rods and bundles to the air in the pool due to pipe rupture.

Section 10
RADIATION PROTECTION

Article 67. Radiation Protection Design

1. Implement regulations to ensure that the radiation dose received by employees at the nuclear power plant is lower than the prescribed dose limits and comply with the ALARA principle.

2. Fully identify all radioactive sources within the nuclear power plant and related radiation risks. Maintain doses from these sources in compliance with the ALARA principle. Ensure the integrity of fuel rod cladding. Control the generation, progression, and impact of corrosion products and activated products.

3. Select materials for structures, systems, and components to minimize material activation potential.

4. Have technical measures to prevent the spread of radioactive substances, radioactive waste, and radioactive contamination within the nuclear power plant.

5. The design of the nuclear power plant must ensure control over employee access to areas with radiation risk and potential radioactive contamination. Use appropriate control measures and ventilation systems to prevent or reduce radiation exposure or contamination.

6. The nuclear power plant must be divided into zones according to radiation levels and contamination levels during its operations (including fuel reloading, maintenance, and inspection); determine potential radiation and contamination levels in case of incidents. Measures must be taken to prevent or reduce radiation exposure.

7. The design of the nuclear power plant must ensure that the radiation dose received by employees during normal operation, fuel reloading, maintenance, and inspection complies with the ALARA principle; use necessary specialized equipment to meet this requirement.

8. Design and install equipment with frequent maintenance schedules or manual operation in low-dose areas to minimize radiation exposure for employees.

9. Provide decontamination facilities for employees and equipment within the nuclear power plant.

Article 68. Radiation Monitoring Measures

1. Equip with radiation monitoring devices during normal operation and in case of incidents in the design. Where feasible, design and install radiation monitoring devices that operate during off-design incidents.

2. Install fixed radiation dose measurement devices to measure local radiation dose rates at locations within the nuclear power plant where employees frequently move and at places with changes in radiation levels during permitted operational states allowing employee access within specified timeframes.

3. Fixed radiation dose measurement devices must have the following features:

a) Display radiation levels at necessary locations within the nuclear power plant during incidents;

b) Provide sufficient information in the control room and control positions to enable employees to take intervention actions when necessary.

4. Fixed radiation monitoring devices must measure radioactivity concentrations in the air in areas where employees regularly work and in places requiring protective measures due to radiation levels. When high radon concentrations are detected, there must be displays in the control room and other necessary locations. Install radiation monitoring devices in areas potentially contaminated by equipment failure or other abnormal situations.

5. Have fixed equipment and laboratories to promptly determine radioactive concentrations in liquid processing systems, gas and liquid samples taken from the nuclear power plant's systems, or from the environment during operational states and incidents.

6. Equip with fixed radiation measurement devices when releasing radioactive exhaust gases or exhaust gases with potential radiation from the nuclear power plant into the environment.

7. Equip with surface contamination radiation measurement devices. Install fixed measuring devices including radiation portal monitors, hand and foot measurement devices at exits from controlled and monitored areas to control radiation for employees and equipment.

8. Establish radiation exposure and radioactive contamination monitoring facilities for employees to evaluate and retain cumulative dose information of employees throughout their working period at the plant.

9. Evaluate radiation exposure and other radiological impacts in the vicinity of the nuclear power plant through environmental dose rate or radioactivity concentration monitoring, paying attention to the following factors:

a) The manner of radiation exposure to the public, including the food chain;

b) Radiological effects (if any) in the local environment;

c) The potential for radioactive accumulation in the environment;

d) The potential for unexpected radioactive dispersion.

Chapter IV
IMPLEMENTING PROVISIONS

Article 69. Effective Date

1. This Circular shall take effect 45 days from the date of signature.

2. During implementation, if there are any difficulties, relevant agencies, organizations, and individuals are requested to promptly report to the Ministry of Science and Technology for review and amendment./.