Circular No. 12/2015/TT-BKHCN on safety analysis for nuclear power plants

CIRCULAR

Regulations on safety analysis for nuclear power plants

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

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

Pursuant to Decree No. 20/2013/NĐ-CP dated February 26, 2013 of the Government stipulating the functions, tasks, powers, and organizational structure of the Ministry of Science and Technology;

At the proposal of the Director of the Radiation and Nuclear Safety Agency and the Head of the Legal Department,

The Minister of Science and Technology issues this Circular prescribing regulations on safety analysis for nuclear power plants.

PART I

GENERAL PROVISIONS

Article 1. Scope of Regulation

This Circular stipulates requirements for safety analysis including deterministic safety analysis and probabilistic safety analysis for nuclear power plants (hereinafter referred to as NPP).

The requirements for safety analysis set forth in this Circular shall be understood as being consistent with the level of detail corresponding to the design approval stages of investment projects, construction permits, operation permits, and during the operation of NPP.

Article 2. Applicability

This Circular applies to organizations and individuals related to the preparation and review of safety analysis reports for NPP.

Article 3. Explanation of Terms

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

1. Deterministic safety analysis is a method of analysis aimed at predicting phenomena occurring after a hypothetical initiating event through the application of specific laws and acceptance criteria. Deterministic safety analysis includes neutron, thermal-hydraulic, radiation, thermo-mechanical, and structural analyses using computational tools.

2. Probabilistic safety analysis is a systematic method of analysis aimed at determining failures and risks with quantified probabilities using computational tools.

3. Sensitivity analysis is an analysis aimed at assessing the degree of change in output results when adjusting input parameters, usually those having the greatest impact.

4. Uncertainty analysis is an analysis aimed at assessing errors in input quantities and computational results.

5. Conservative approach is the use of computational programs, models, input data, and assumptions to assess safety in the most cautious manner.

6. Safety margin is the level of protection between the analysis and evaluation results of parameters related to safety and the safety limits prescribed by the competent authority.

7. Verification is the process of determining the correctness of methods, programs, or models according to descriptions, plans, or requirements.

8. Validation is the process of determining the experimental suitability of methods, programs, or models according to their functions.

9. Hypothetical initiating event is a hypothetical event arising directly from structural, system, component failure, operational error, or internal and external hazards when the NPP operates at rated power, low power, or shutdown state.

10. Departure from nucleate boiling ratio (DNBR) is the ratio of the limiting heat flux to the actual local heat flux of fuel rods.

Chapter II

GENERAL REQUIREMENTS FOR SAFETY ANALYSIS METHODS OF NUCLEAR POWER PLANTS

ANALYSIS OF THE NUCLEAR POWER PLANT SAFETY

Article 4. Scope of safety analysis application

1. Safety analysis of NPP must be conducted using both deterministic and probabilistic safety analysis methods to evaluate the safety level of the plant under different operating states and conditions.

2. Safety analysis of NPP must be carried out in all states, including normal operation, abnormal states, design-basis accidents, beyond-design-basis accidents, and severe accidents.

3. Safety analysis of NPP must determine the frequency of occurrence of hypothetical initiating events, physical and thermal-hydraulic parameters of important safety systems, the status of physical barriers, and the consequences of radioactive leakage into the environment.

4. Safety analysis must include events arising from internal and external hazards and processes that may damage containment barriers or increase the risk of radioactive leakage into the environment. External hazards with low frequencies but capable of leading to core melt must be considered in severe accident analysis.

5. Selection of events for analysis and subsequent developments must be based on a systematic and logical approach. Sufficient justification must be provided for identifying all relevant accident scenarios.

6. Specific requirements for the scope of probabilistic safety analysis implementation, including:

a) Conducting Level 1 probabilistic safety analysis to determine the frequency of events leading to core melt; estimating core melt frequency; evaluating strengths and weaknesses of safety systems and operational procedures to prevent core melt;

b) Conducting Level 2 probabilistic safety analysis to identify pathways for radioactive release in severe accidents, estimate the extent and frequency of releases; evaluate the adequacy of preventive and mitigative measures to reduce radioactive dispersion into the environment;

c) Performing analysis for reactors and spent fuel pools in all operating modes and states of NPP;

d) Initiating event analysis including internal events within the plant, human errors, internal and external hazards.

7. The scope and level of detail of safety analysis must correspond to the radiological consequences and frequency of events.

Article 5. Safety Analysis in Nuclear Power Plant Design

1. Determining the design basis for important safety components; their role in reducing hypothetical initiating events as well as in the event sequence.

2. Safety analysis must demonstrate that the design meets the required depth of protection levels.

3. Safety analysis must justify the application of assumptions, methods, uncertainties, and conservatism in the design.

4. Applying an appropriate quality assurance program in conducting safety analysis.

Article 6. Results of Deterministic Safety Analysis and Probabilistic Safety Analysis

1. The results of deterministic safety analysis include comparing the analysis outcomes with acceptance criteria stipulated in Chapter III of this Circular and the following contents:

a) Confirming the suitability of the design basis for all important safety components; the suitability of operational limits, conditions, and necessary operator actions;

b) Confirming that hypothetical initiating events are consistent with the characteristics of the site and nuclear power plant design;

c) Justifying the management of abnormal states and accidents within the design basis through activation of safety systems is consistent with acceptance criteria;

d) Justifying the management of beyond-design accidents using safety features without being affected by accident sequences is consistent with acceptance criteria.

2. The results of probabilistic safety analysis include comparing the analysis outcomes with acceptance criteria stipulated in Chapter III of this Circular and the following contents:

a) Justifying that no hypothetical initiating event contributes excessively to the overall risk or significantly to the uncertainty of the analysis results;

b) Justifying that layers of defense in depth must be maximally independent to the extent practically achievable;

c) Justifying the prevention of sudden adverse effects on safety due to fluctuations in input parameters.

Article 7. Specific Requirements for Deterministic Safety Analysis

1. When performing deterministic safety analysis, it must ensure sufficient safety margin between the calculated values of important parameters and threshold values leading to radioactive release even when using the best estimation method.

2. Deterministic safety analysis for design purposes must ensure conservatism, including reasonable consideration of model uncertainties, except for analysis of beyond-design accidents.

3. Selection of computational data and assumptions must consider the uncertainties of the following factors:

a) Initial operating conditions of the plant;

b) Operational capability of safety systems;

c) Operator actions;

d) Grid availability to start up safety systems.

4. For each event analysis, the following provisions must be implemented:

a) Determine relevant acceptance criteria and physical limiting parameters;

b) Select conservative initial and boundary conditions corresponding to each acceptance criterion. Sensitivity analysis must be performed to justify these selections if necessary;

c) For systems and equipment not designed to operate under certain states, they must be assumed to be failed or operated in a manner exacerbating the initiating event, unless their operational reliability can be proven high;

d) In analyzing design-basis accidents, the most severe single failure must be assumed to occur during safety system operation.

Article 8. Uncertainty Analysis and Sensitivity Analysis

1. The results of safety analysis must include the results of uncertainty analysis and sensitivity analysis.

2. Uncertainty must be quantified and clearly explained with regard to the method of handling, taking into account the origin and nature of uncertain factors.

Article 9. Calculation Programs in Safety Analysis

1. Priority should be given to using the best estimation calculation program capable of simulating the most realistic important phenomena and operations of systems within the plant.

2. Calculation methods, calculation programs, and calculation models in safety analysis must be verified and validated.

Article 10. Utilization of Operating Experience in Safety Analysis

1. Collecting and evaluating data on operational activities that can be used for safety analysis, including reviewing abnormal conditions and incidents that have occurred during operation at similar nuclear power plants.

2. Data on operating experience includes:

a) Records of abnormal conditions and incidents occurring during operation;

b) Operator errors;

c) Operation of safety systems;

d) Reliability of important safety components;

đ) Radiation dose;

e) Generation of radioactive waste.

3. Throughout the operational period of the nuclear power plant, data must be collected based on the plant's safety quality indicators. Operating experience data must be utilized appropriately, along with advancements in scientific and technical methods of safety analysis and related research results to update safety analysis outcomes and evaluate management systems.

Chapter III

ACCEPTANCE CRITERIA FOR SAFETY ANALYSIS

Article 11. Requirements for Establishing Acceptance Criteria

1. Acceptance criteria must be established for all operational states and accident conditions. These criteria must ensure sufficient depth of protection, ensuring that unacceptable harm to humans and the environment does not occur.

2. Acceptance criteria are established based on the following factors:

a) Characteristics of the assumed initiating event, particularly its frequency;

b) Type of reactor technology;

c) Actual conditions of the plant, particularly self-sufficiency in electricity;

d) Scope and conditions of application of each criterion. More stringent criteria must be applied to events with higher frequencies.

3. Acceptance criteria for each operational state or accident condition include:

a) General criteria related to radiological release consequences;

b) Specific criteria related to the integrity of radiation release prevention layers, including fuel pellets, fuel rod cladding, reactor coolant pressure boundary, and reactor containment.

Article 12. Acceptance Criteria for Deterministic Safety Analysis of Radiological Release Consequences

1. Radiation doses for plant personnel and the public at operational states throughout the plant's lifetime must comply with the ALARA principle. The dose limit for an individual must be less than 1 mSv/year.

2. All incidents within the design basis must not cause significant radiological impact at the site or off-site and do not require off-site protective measures. The dose limit for the public due to incidents within the design basis (excluding ingestion of radioactive substances through food) must be less than 5 mSv/year.

3. Incidents outside the design basis that may lead to large-scale radioactive releases into the environment must be practically excluded. For other incidents outside the design basis, measures to limit release over a certain period and within a defined range must be implemented to provide sufficient time to deploy protective measures for the public.

4. Radioactive material release from severe accidents must not result in the following consequences:

a) Acute radiation injury to the health of nearby residents;

b) Long-term restrictions on land and water use over a wide area;

c) Release of Cs-137 into the environment exceeding 30 TBq;

d) Three months after the incident, the combined activity of other radionuclides deposited, excluding cesium, must not exceed the harmful effects of cesium release as specified in point c of this clause.

Article 13. Common criteria for deterministic safety analysis

1. The characteristics of design, the automatic activation capability of the safety system combined with the necessary actions of the operating staff must be sufficiently effective to ensure:

a) Not worsening the state of the plant;

b) Not causing additional single failures;

c) Not losing the operational capability of the safety system or the essential safety features needed to mitigate the consequences of an accident.

2. Systems used to mitigate the consequences of events must have the capacity to withstand maximum loads, stresses, and environmental conditions corresponding to the analyzed event.

3. Pressure in primary and secondary systems must not exceed the design limit for each specific state of the plant.

4. The assumed initiating event may lead to damage to the fuel rod cladding but must still meet the acceptance criteria for radiological release consequences as stipulated in Article 12 of this Circular.

5. For all accidents within the design basis, the geometry of the core must be maintained to ensure adequate core cooling.

6. Accidents within the design basis must not cause temperatures, pressures, pressure differentials, or other loads acting on the reactor vessel exceeding the values used for design basis.

Article 14. Specific acceptance criteria for deterministic safety analysis

1. Determine specific acceptance criteria to ensure sufficient safety margin and integrity of the radioactive release barrier layers. Consider applying the following specific acceptance criteria:

a) Criteria related to the integrity of the fuel pellet structure including: maximum fuel pellet temperature, maximum value of average fuel enthalpy by radius (along with their dependence on burn-up depth and fuel composition);

b) Criteria related to the integrity of the fuel rod cladding including: minimum departure from nucleate boiling ratio (DNBR) (for pressurized water reactors) or power-to-flow ratio (CPR) (for boiling water reactors), maximum fuel rod cladding temperature, maximum local oxidation level of fuel rod cladding;

c) Criteria related to the integrity of the reactor core region including: subcriticality level, maximum hydrogen generation due to fuel rod cladding oxidation, maximum number of damaged fuel rods and fuel assemblies;

d) Acceptance criteria related to the integrity of the reactor coolant system (primary loop) including: maximum coolant temperature and pressure; temperature, pressure, and stress changes caused at the reactor coolant system pressure boundary; no plug, brittle fracture, or ductile failure due to assumed defects in the reactor vessel;

e) Acceptance criteria related to the integrity of the secondary loop including: maximum temperature and pressure, pressure and temperature changes of the working fluid in secondary loop equipment;

f) Acceptance criteria related to the integrity of the reactor containment and limiting radioactive release to the environment including: maximum and minimum containment internal pressure and temperature, maximum pressure differential between two containment walls (in case of double-wall containment design), leakage rate, flammable gas concentration, environmental conditions acceptable for system operation parameters;

2. For assumed initiating events occurring during shutdown or leading to degradation of protective layer integrity, stricter acceptance criteria must be applied to prevent fuel assembly exposure.

Article 15. Criteria for acceptance of probabilistic safety analysis

Clause 1. For Level 1 probabilistic safety analysis, the total frequency of core melt region occurrence must be less than 10/lò.năm.^-5/reactor.year.

Clause 2. For Level 2 probabilistic safety analysis, the total frequency of radioactive Cs-137 release exceeding 30 TBq into the environment must be less than 10^-6/reactor.year.

Chapter IV

IMPLEMENTING PROVISIONS

Article 16. Transitional Provisions

In cases where organizations or individuals submit a probabilistic safety analysis report to the Ministry of Science and Technology before this Circular takes effect, they are not required to comply with the provisions of this Circular.

Article 17. Effective Date

Clause 1. This Circular takes effect from September 5, 2015.

Clause 2. During implementation, if there are any difficulties or new issues arise, it is requested that agencies, organizations, and individuals promptly reflect these to the Ministry of Science and Technology for consideration of amendments and supplements./.