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MINISTRY OF SCIENCE AND TECHNOLOGY OF VIETNAM
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SOCIALIST REPUBLIC OF VIETNAM
Independence - Freedom - Happiness
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No. 07/2026/TT-BKHCN
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Hanoi, March 31, 2026
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CIRCULAR
NUCLEAR SAFETY FOR SITE, DESIGN,
CONSTRUCTION, AND SAFETY ANALYSIS OF NUCLEAR POWER PLANTS
Pursuant to the Law on Atomic Energy dated June 27, 2025;
Pursuant to Decree No. 55/2025/ND-CP dated March 2, 2025 of the
Government of Vietnam on functions, tasks, entitlements, and organizational
structure of the Ministry of Science and Technology of Vietnam;
Pursuant to Decree No. 316/2025/ND-CP dated December 10, 2025 of the
Government of Vietnam elaborating certain articles and implementation measures
of the Law on Atomic Energy regarding nuclear power plants and research
reactors;
At the request of the Director of the Vietnam Agency for Radiation and
Nuclear Safety and the Director of the Department of Legal Affairs;
The Minister of Science and Technology of Vietnam hereby promulgates the
Circular on nuclear safety for the site, design, construction, and safety
analysis of nuclear power plants.
Chapter I
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Article 1. Scope
This
Circular elaborates on certain articles of Decree No. 316/2025/ND-CP dated
December 10, 2025 of the Government of Vietnam elaborating certain articles and
implementation measures of the Law on Atomic Energy regarding nuclear power
plants and research reactors, including:
1.
Site survey and site assessment for the construction of nuclear power plants,
excluding mobile nuclear power plants, as prescribed in Clause 2 Article 14.
2.
The design of water-cooled reactor nuclear power plants, as prescribed in
Clause 5 Article 8; defense in depth, as prescribed in Clause 4 Article 4.
3.
Construction of nuclear power plants as prescribed in Clause 6 Article 24.
4.
Safety analysis for nuclear power plants as prescribed in Clause 7 Article 9.
5.
The quality management system applicable to site survey, site assessment, design,
construction, trial operation, operation, and operational termination of
nuclear power plants, as prescribed in Clause 4 Article 10.
Article 2. Regulated entities
This
Circular applies to domestic organizations and individuals, overseas
Vietnamese, foreign organizations and individuals, and international
organizations conducting activities related to nuclear power plants within the
territory of Vietnam as prescribed in Article 1 of this Circular.
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For
the purpose of this Circular, the following terms shall be construed as follows:
1.
Component refers to equipment or an independent part, or equipment or
parts constituting a system.
2.
Structure refers to construction works that serve a shielding,
protective, or supporting function, such as a building, a spent nuclear fuel
storage pool, a support pedestal, or a suspension frame.
3.
Operating modes include start-up, operation at rated power, operation at
partial rated power, reactor shutdown, and fuel replacement and fuel shuffling.
4.
Design basis refers to the set of conditions, processes, and factors
arising from natural or human-induced causes that are taken into account in the
design of a nuclear power plant, such that, when such conditions, processes, or
factors occur, the safety systems continue to operate as designed and the
safety limits approved by the competent authority remain ensured.
5.
Redundancy refers to the provision of structures, systems, and
components (hereinafter collectively referred to as “items”) capable of
substituting for one another to independently perform the same function,
regardless of the operating status or failure of any item.
6.
Diversity refers to the provision of two or more redundant systems or
components having different attributes to perform the same function, thereby
minimizing the likelihood of common-cause failures.
7.
Independent assessment is an assessment conducted by an organization or
individual not involved in the process being assessed to ensure objectivity and
enhance the effectiveness of safety assurance.
8.
Construction unit refers to the unit responsible for managing
construction activities, including civil works, architectural works,
fabrication, installation, and testing of items important to safety at a
nuclear power plant.
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10.
Site refers to the geographical area proposed for the construction of a
nuclear power plant.
11.
Site vicinity refers to the geographical area with a radius of at least
5 km from the proposed nuclear power plant site.
12.
Subregion refers to the geographical area with a radius of at least 25
km from the proposed nuclear power plant site.
13.
Region refers to the geographical area with a radius of approximately
300 km from the proposed nuclear power plant site.
14.
Design extension conditions are postulated accident conditions that are
not considered in accidents concerning the design basis, but are taken into
account in the design of a nuclear power plant using the best-estimate
approach, under which releases of radioactive materials are maintained within
acceptable limits. Design extension conditions include conditions that
do not result in significant fuel damage and conditions leading to core melting.
15.
Accident conditions refer to events deviating from normal operation that
have a very low frequency of occurrence and are more severe than anticipated
operational occurrences. Accident conditions include design basis accidents and
design extension conditions.
16.
Safety margin refers to the margin between the results of the analysis
and assessment of safety-related parameters and the safety limits prescribed by
the competent authority.
17.
Item important to safety refers to an item belonging to a safety class,
or an item whose malfunction or failure may result in radiation exposure to
workers or the public.
18.
Cliff-edge effect refers to the phenomenon whereby severe abnormal
operating conditions of an establishment arise due to an abrupt transition from
one state to another as a result of only a small variation in a parameter or
input value.
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20.
Safety system refers to a system provided to ensure the safe shutdown of
the reactor, the removal of residual heat from the reactor core, and the
limitation of the consequences of anticipated operational occurrences and
design basis accidents.
21.
Safety system support system refers to a collection of equipment providing
cooling, lubrication, and power supply services to the protection system and
the safety actuation system.
22.
Quality management system refers to the set of policies, procedures,
resources, and responsibilities established to ensure that nuclear safety,
radiation safety, and nuclear security requirements are complied with
throughout the stages of a nuclear power plant.
23.
Karst refers to the combination of geological processes and phenomena
occurring on or beneath the ground surface, primarily due to the chemical
dissolution of rocks and soils, resulting in the formation of cavities, as well
as causing the degradation and alteration of geological structures, ground
conditions, groundwater mechanisms, topographical characteristics, and
hydrological network systems.
24.
Practical elimination refers to the requirement that design, analysis,
and assessment be performed such that the likelihood of occurrence of a
particular state or event is demonstrated to be extremely unlikely, thereby
reasonably allowing it to be considered eliminated for all practical purposes.
25.
Ultimate heat sink refers to a system, structure, component, or
environmental medium to which heat can always be transferred.
26.
ALARA principle refers to the application of technical and
administrative measures to restrict individual radiation doses to radiation
workers and the public, and to keep the number of exposed individuals as low as
is reasonably achievable, taking into account economic and social factors.
27.
Active volcano refers to a volcano exhibiting reliable indications of
its potential to generate hazardous phenomena, including non-eruptive phenomena
that may affect the site during the lifecycle of the nuclear power plant.
28.
Physical separation refers to separation by distance, orientation,
suitable barriers, or a combination of these methods.
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30.
Probabilistic safety analysis refers to a systematic analytical method
for identifying failures and risks, with probabilities of occurrence quantified
using computational tools.
31.
Sensitivity analysis refers to an analysis conducted to assess the
degree of variation in output results resulting from adjustments to input
parameters, typically those having the greatest influence.
32.
Uncertainty analysis refers to an analysis conducted to assess errors
associated with input quantities and calculation results.
33.
Graded approach refers to an approach in which the stringency of control
measures and the applicable conditions are determined in proportion to the
probability of occurrence, the severity of consequences, and the level of risk.
34.
Ageing management refers to the set of technical, operational, and
maintenance activities carried out to control, within acceptable limits, the
ageing degradation of structures, systems, and components.
35.
Accident management refers to the set of actions implemented during the
progression of an accident to prevent its escalation into a severe accident,
mitigate the consequences of a severe accident, and achieve a stable and
long-term safe state.
36.
Single failure refers to the failure of a system or component to perform
its designed safety function, or any consequential failure resulting from the
loss of that safety function.
37.
Common-cause failure refers to the failure of two or more structures,
systems, or components due to the same specific event or cause.
38.
Design basis accident refers to a postulated accident for which a
nuclear power plant or research reactor is designed to withstand in accordance
with established design criteria using a conservative approach, under which the
release of radioactive materials is controlled within the prescribed
permissible limits.
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40.
Postulated initiating event refers to a postulated event identified in
the design that may lead to an anticipated operational occurrence or an
accident condition.
41.
Single failure criterion refers to the criterion applied to a system to
ensure that it remains capable of performing its required function in the event
of a single failure.
42.
Safe state refers to the state of a nuclear power plant following an
anticipated operational occurrence or an accident condition, in which the
reactor remains subcritical, and the fundamental safety functions are ensured
and maintained in a stable condition over the long term.
43.
Controlled state refers to the state of a nuclear power plant following
an anticipated operational occurrence or an accident condition, in which the
fundamental safety functions are ensured and maintained for a sufficient period
to enable measures to be taken to achieve the safe state.
44.
Plant state refers to all possible states of a nuclear power plant,
including normal operation and anticipated operational occurrences
(collectively referred to as “operating states”), as well as states involving
design basis accidents and design extension conditions, including accidents
involving core melting.
45.
Conservative approach refers to the use of computational codes, models,
input data, and assumptions to provide the most conservative assessment of
safety.
46.
Safety feature refers to the design characteristics provided to prevent
accidents or mitigate their consequences.
47.
Anticipated operational occurrence refers to an operational deviation from
normal operation that is expected to occur at least once during the operating
lifetime of a nuclear power plant, but which does not cause significant damage
to items important to safety or lead to an accident condition due to
appropriate design measures.
48.
Site design parameter refers to a parameter required for the design of a
nuclear power plant. The design of a nuclear power plant is considered suitable
for the site if the parameters used in the plant design envelop the
corresponding site design parameters.
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50.
Subsurface materials refer to materials beneath the ground surface,
including soils, rocks, sediments, geological structures, and hydrogeological
components that may affect the safety of a nuclear power plant.
51.
Construction of a nuclear power plant refers to the process of fabricating
and installing the components of a nuclear power plant, carrying out civil
construction works, installing components and equipment, and conducting tests
to ensure that the structures, systems, and components are constructed,
fabricated, and installed in accordance with the approved design.
52.
Validation refers to the process of determining the conformity of a
method, computational code, or computational model with experimental evidence
for its intended functions.
53.
Verification refers to the process of determining the correctness of a
method, computational code, or computational model against its description,
intended purpose, or specified requirements.
54.
Hazard refers to a natural or human-induced phenomenon or process that
may give rise to an accident affecting a nuclear power plant.
Chapter II
NUCLEAR SAFETY
REQUIREMENTS FOR NUCLEAR POWER PLANT SITE SURVEY AND SITE ASSESSMENT
Section 1. GENERAL REQUIREMENTS FOR SITE SURVEY AND SITE ASSESSMENT
Article 4. Requirements for site survey and site assessment
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a)
Identification of candidate sites;
b)
Preliminary assessment of site suitability at the stage of approval of
investment guidelines, in accordance with Article 13 of Decree No.
316/2025/ND-CP and other relevant laws;
c)
At the site approval stage, determining the absence of indications of exclusion
factors, assessing adverse factors, and verifying the suitability of the site
for the construction of a nuclear power plant. Site survey, site assessment,
and determination of site design parameters shall be carried out in accordance
with Sections 1, 2, and 3 of Chapter II of this Circular and other relevant
laws;
d)
During the project implementation stage and the operation of the nuclear power
plant, determining the site design parameters, establishing the design basis,
and conducting periodic monitoring and assessment of the site. Periodic
monitoring and assessment of external hazards and site conditions shall be
conducted in accordance with Section 4 of Chapter II of this Circular.
2.
A site shall be considered suitable for the construction of a nuclear power
plant if there are no exclusion factors, measures are available to mitigate
adverse factors, and a feasible and effective emergency preparedness and
response plan is in place.
3.
Site suitability shall be assessed on the basis of relevant data and
methodologies. Conservative criteria shall be established in accordance with
site-specific accident scenarios and shall be consistent with the overall
suitability of the site.
4.
The determination of site suitability shall be based on the characteristics of
the nuclear power plant, including its intended operation, the quantity and
nature of potential radioactive release sources, and their impacts on humans
and the environment.
5.
In the overall assessment of site suitability, other characteristics shall also
be taken into account, including the availability of cooling water, extreme
environmental conditions, and their potential impacts on the safe and
continuous operation of the nuclear power plant.
6.
Exclusion factors include:
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b)
The possibility of earthquake-induced ground motion that could endanger the
nuclear power plant, where no mitigation measures are available.
7.
Adverse factors include:
a)
The presence of an active fault outside the site vicinity that could endanger the
nuclear power plant;
b)
Evidence of seismic activity within the study area that could endanger the
nuclear power plant;
c)
Evidence of an active volcano within the study area that could endanger the
nuclear power plant;
d)
Developing karst processes and phenomena at the site that could endanger the
nuclear power plant;
dd)
Bedrock intended for construction of the reactor building that is highly
fractured, severely cracked, or weathered;
e)
Foundation ground intended for construction of other facilities of the nuclear
power plant that consists of weak soil or soil susceptible to liquefaction,
amplification of ground motion, swelling, or significant shrinkage;
g)
Meteorological phenomena such as typhoons, tornadoes, or strong winds that
could endanger the nuclear power plant;
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i)
A distance of less than 4 km between the site and an aircraft take-off and
landing corridor;
k)
A distance of less than 10 km between the site and an airport with a runway
longer than 1800 m and an average of not more than 5000 flights per week; or a
distance of less than 16 km for an airport with a runway longer than 1800 m and
an average exceeding 5000 flights per week;
l)
Insufficient cooling water resources for the safe operation of the nuclear
power plant;
m)
Other adverse factors that could endanger the nuclear power plant.
8.
Indications of exclusion factors and adverse factors shall be collected and
derived from prehistoric and historical information, official data provided by
competent authorities and organizations, and other available survey,
monitoring, and assessment data.
9.
Site assessment shall be fully documented in accordance with the requirements
of the quality management system.
Article 5. Scope of site survey and site assessment
1.
The scope of site survey and site assessment shall include factors related to
the site and factors related to the interaction between the site and the
nuclear power plant, for all operating states and accident conditions.
2.
The scope of site survey and site assessment shall include all natural and
human-induced external hazards, monitoring activities, and site-specific
parameters relevant to the safety of the nuclear power plant. The scope and
level of detail of the site assessment process shall be determined in
accordance with the graded approach.
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a)
The effects of natural and human-induced external hazards that may affect the
site;
b)
Characteristics of the site and its surrounding environment that may affect the
dispersion of radioactive materials released from the nuclear power plant, resulting
in harm to humans and the environment;
c)
Population density and distribution, and other external characteristics that
may affect the feasibility and effectiveness of the emergency preparedness and
response plan, as well as the assessment of risks to individuals and the
community.
4.
Geological and tectonic surveys, assessments, and mapping of the area
surrounding the site shall be conducted at a scale of 1:500.000 within a radius
of 150 km; at scales from 1:50.000 to 1:25.000 within a radius of 25 km; and at
a scale of 1:10.000 or larger within a radius of 5 km from the site.
5.
The application of nuclear safety requirements for nuclear power plant site
survey and site assessment shall be commensurate with the potential hazards
associated with the nuclear power plant.
Article 6. Characteristics of the site and the area surrounding the site
1.
Studies shall be conducted on the characteristics of the site and the area
surrounding the site that may adversely affect the safety of the nuclear power
plant, as well as the potential radiological impacts of the nuclear power plant
on humans and the environment.
2.
Natural phenomena and human activities in the area surrounding the site that
may adversely affect the safety of the nuclear power plant shall be identified
and assessed. The level of assessment shall be commensurate with the
significance to safety of the potential hazards at the site.
3.
The characteristics of the natural environment in the area surrounding the site
that may be affected by the potential radiological impacts of the nuclear power
plant shall be studied and assessed for all operating states, accident
conditions, and all stages of the lifecycle of the nuclear power plant.
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5.
The site and the area surrounding the site shall be investigated to assess
existing and future characteristics that may adversely affect the safety of the
nuclear power plant.
Article 7. Identification of site-specific hazards
1.
Potential natural and human-induced external hazards, as well as human
activities that may affect the area surrounding the site, shall be identified
through a screening process.
2.
The procedures and criteria used in the hazard screening process for the site
shall be justified and documented.
3.
During the screening process, the scope of assessment of external events shall
encompass the severity and frequency of occurrence of such events throughout
the design and safety analysis of the nuclear power plant, including events
with a low probability of occurrence but severe consequences that may
contribute to the overall level of risk.
4.
External hazards that are not screened out shall be assessed and subsequently
used to establish the site design parameters according to the significance of
their impacts on the safety of the nuclear power plant.
Article 8. Assessment of natural and human-induced external hazards
1.
The impacts of natural and human-induced external hazards on the safety of the
nuclear power plant shall be assessed throughout the lifecycle of the nuclear
power plant.
2.
In the site assessment, consideration shall be given to the frequency and
severity of natural and human-induced external events (including combined
events) that may adversely affect the safety of the nuclear power plant.
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4.
Appropriate methods and models shall be used, where necessary, for hazard
analysis during the site assessment and the design of the nuclear power plant.
Uncertainty analysis of the methods and input data shall be performed as part
of the hazard assessment.
5.
When conducting the assessment of external hazards, consideration shall be
given to the possibility of accidents occurring simultaneously or in close
succession, and the interrelationship and mutual impacts among such accidents
shall be analyzed.
6.
The results of the hazard assessment shall be used as input for establishing
the site design parameters.
7.
During the site assessment, potential accidents such as explosions, chemical
releases, and thermal releases that may affect the safety of the nuclear power
plant or result in the release of radioactive materials shall be considered.
Potential accidents associated with interactions between radioactive and
non-radioactive substances shall also be taken into account.
Article 9. Site protection measures
1.
The need for site protection measures shall be assessed where the design of the
nuclear power plant is incapable of withstanding the impacts of natural and
human-induced external hazards, taking into account whether an adequate safety
margin is available.
2.
The availability and feasibility of technical solutions for implementing site
protection measures shall be assessed. A site shall be considered unsuitable if
such technical solutions are unavailable.
3.
Where site protection measures are required, the assessment shall take into
account the uncertainties associated with the extreme values of the parameters
used to characterize the severity of natural and human-induced external
hazards. Site protection measures shall be classified, designed, constructed,
maintained, and operated in accordance with the significance to safety of the
impacts that such hazards may cause.
Article 10. Site Assessment for multiple nuclear power units
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2.
The possibility that a hazard originating from one nuclear power unit may
adversely affect another nuclear power unit located at the same site or within
the same subregion shall be assessed.
3.
The cumulative impacts of accidents occurring at nuclear power units located at
the same site or within the same subregion on humans and the environment shall
be assessed.
Article 11. Changes in hazards and site characteristics over time
1.
External hazards and site characteristics that may change over time shall be
assessed, including the impacts of climate change such as increasing
temperatures, sea-level rise, increased frequency and intensity of hazardous
meteorological phenomena, and changes in precipitation. The potential impacts
of such changes throughout the operating lifecycle of the nuclear power plant
shall be assessed.
2.
Uncertainties in forecasting potential changes in external hazards and site
characteristics over time shall be taken into account by applying appropriate
safety margins to the relevant site design parameters.
Article 12. Consideration of ultimate heat sink
1.
Consideration shall be given to natural and human-induced external hazards that
may affect the feasibility and reliability of the ultimate heat sink.
2.
Consideration of the ultimate heat sink shall be based on the following data:
a)
Air temperature and humidity;
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c)
Water quality characteristics, including turbidity, suspended solids content,
floating debris, and changes in the chemical and biochemical properties of the
water, including both natural and human-induced changes;
d)
The availability and stability of the water flow rate, in the case of a river,
including the minimum and maximum water levels and the period during which the
cooling water source satisfies the minimum safety requirements, taking into
account the risk of failure of water control structures.
3.
All natural and human-induced external hazards that may result in the loss of
the ultimate heat sink shall be identified and assessed.
Article 13. Potential impacts of nuclear power plant on humans and
environment
1.
In determining the potential radiological impacts of a nuclear power plant on
the surrounding area under all operating states and accident conditions,
including accidents requiring the initiation of emergency response actions,
appropriate estimates shall be made of the potential release of radioactive
materials, taking into account the design and safety features of the nuclear power
plant.
2.
The potential impacts of the nuclear power plant on humans and the environment
shall be estimated by considering postulated accident scenarios and taking into
account the feasibility of effective emergency response actions at the site and
in the off-site area. These estimates shall be confirmed once the design and
safety features of the nuclear power plant have been established.
3.
Direct and indirect pathways through which radioactive materials released from
the nuclear power plant may adversely affect the public and the environment
shall be identified and assessed, taking into account the characteristics of
the site and the area surrounding the site, including population distribution,
with particular attention paid to the transport and accumulation of radioactive
materials in the biosphere.
4.
It shall be demonstrated that the information used to assess potential impacts
on the population under accident conditions complies with the requirements and
acceptance criteria applicable to the site assessment.
5.
The impacts of ionizing radiation on the public shall be assessed.
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a)
The exclusion area is the area surrounding the nuclear power plant site and
shall simultaneously satisfy the following requirements:
The
effective dose from the plant to an individual located at the outer boundary of
the exclusion area shall not exceed 0,25 mSv per year under normal operation;
The
effective dose to an individual located at the outer boundary of the exclusion
area shall not exceed 250 mSv, and the equivalent dose to the thyroid, due to
I-131, shall not exceed 3000 mSv, during any 2-hour period following the occurrence
of a postulated accident.
b)
The low population zone is the area surrounding the exclusion area, at the
outer boundary of which an individual shall receive a total effective dose not
exceeding 250 mSv or an equivalent dose to the thyroid due to I-131 not
exceeding 3000 mSv during the passage of the postulated radioactive cloud.
Article 14. Assessment of feasibility of effective emergency response
actions
1.
The feasibility of effective emergency response actions at the site and in the
off-site area shall be assessed, taking into account the characteristics of the
site and the off-site area.
2.
Consideration shall be given to nuclear power units located at the same site
and within the same subregion, with particular attention paid to nuclear power
units at which accidents may occur simultaneously.
3.
The relationship between external hazards and infrastructure conditions at the
site and in the off-site area shall be considered when assessing the
feasibility of effective emergency response actions.
Article 15. Data collection during site assessment
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2.
Data shall be collected on natural and human-induced external hazards that may
adversely affect the safety of the nuclear power plant throughout its operating
lifecycle.
3.
The data collected shall include the following principal information:
a)
Information on natural and human-induced external hazards, including
information on hazard sources, hazard propagation, and potential impacts on the
nuclear power plant, humans, and the environment;
b)
Information describing the conditions of the site and the area surrounding the
site;
c)
Information on technical and administrative measures for protecting the site,
and on mitigation measures;
d)
Information on the potential impacts of the nuclear power plant on humans and
the environment under operating states and accident conditions;
dd)
Information necessary for effective on-site and off-site emergency response
actions under all environmental conditions and for all plant states;
e)
Information on access conditions to the site.
4.
Information and records, if any, concerning the occurrence and severity of
significant natural phenomena during prehistoric, historical, and recent
periods shall be collected. The reliability, accuracy, temporal and spatial
relevance, and completeness of such information and records shall be analyzed.
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6.
Data collected for the site investigation shall be sufficient in quantity and
meet the quality requirements necessary to support the selected hazard
assessment methodology.
7.
The level of detail of the information collected for each hazard shall be
commensurate with the distance from the hazard source to the site and the
magnitude of its potential impacts on the site. Data collection shall be fully
documented, including the associated uncertainties.
Section 2. REQUIREMENTS FOR SURVEY AND ASSESSMENT OF NATURAL EXTERNAL
HAZARDS
Article 16. Assessment of capability of fault to be active
1.
An active fault is a tectonic fault capable of causing displacement at or near
the ground surface. On the basis of geological, geophysical, geodetic, and
seismic data, a fault shall be considered an active fault where it has any of
the following characteristics:
a)
There is evidence of displacement at or near the ground surface occurring at
least once, or repeatedly in a cyclic manner, within approximately the past 130
thousand years in areas of high tectonic activity, including plate boundaries,
or within approximately the past 2,58 million years in areas of lower tectonic
activity, including intraplate areas;
b)
It has a structural relationship with another known active fault, and it has
been demonstrated that displacement of that fault may cause displacement at or
near the ground surface of the other fault;
c)
There is a reasonable basis for concluding that displacement at or near the
ground surface may occur under the present-day tectonic environment at the
nuclear power plant site, on the basis that the predicted maximum earthquake
associated with the seismogenic structure is sufficiently large and occurs at
an appropriate depth.
2.
Large-scale geological faults located within a specified distance from the site
and having significant implications for the safety of the nuclear power plant
shall be assessed to determine whether such faults are to be considered active
faults. Methods and investigations shall be sufficiently detailed to support
safety-related decisions. For an active fault, the potential impacts on the
safety of the nuclear power plant in terms of ground displacement or fault
displacement hazards shall be assessed.
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Article 17. Assessment of ground motion hazards
1.
Ground motion hazards shall be assessed to provide:
a)
The input data necessary for the seismic design or safety upgrading of the
structures, systems, and components of the nuclear power plant;
b)
Input data for conducting the deterministic safety analysis or probabilistic
safety analysis required throughout the operating lifecycle of the nuclear
power plant.
2.
Earthquake-induced ground motion hazards shall be assessed using an appropriate
method. The impacts of ground motion in combination with the impacts of other
earthquake-induced events shall be considered. The possibility of human-induced
earthquakes shall also be considered.
Article 18. Assessment of volcanic hazards
1.
Hazards arising from volcanic activity that may adversely affect the safety of
the nuclear power plant shall be assessed. The impacts of volcanic phenomena in
combination with other volcanic hazards, including volcanic ejecta fall, shall
be considered.
2.
Volcanoes capable of becoming active shall be identified and assessed. The
volcanic characteristics of the area surrounding the site shall be considered
to ensure that potentially hazardous volcanic phenomena have been taken into
account.
3.
Hazards posed by volcanoes capable of eruption shall be assessed to provide the
input information necessary for determining the site design parameters and for
conducting the deterministic safety analysis and probabilistic safety analysis
throughout the operating lifecycle of the nuclear power plant.
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Article 19. Assessment of hazardous meteorological phenomena
1.
Hazardous meteorological phenomena and their possible combinations that may
adversely affect the safety of the nuclear power plant shall be assessed.
2.
Meteorological phenomena, including wind, precipitation, air and water
temperatures, humidity, storm surges, and sandstorms or dust storms, shall be
assessed. Combined meteorological phenomena shall be assessed with respect to
extreme values of meteorological parameters on the basis of available data.
3.
Appropriate methods shall be applied to assess meteorological hazards, taking
into account available data, including measured and historical data, as well as
past changes in the characteristics of the area surrounding the site.
Article 20. Assessment of hazardous meteorological events
1.
The likelihood, frequency, and severity of hazardous meteorological events
shall be assessed.
2.
The likelihood, frequency, and severity of lightning in the site vicinity shall
be assessed.
3.
The likelihood, frequency, and severity of tornadoes, storms, and associated
flying objects shall be assessed. Hazards associated with tornadoes and storms
shall be characterized by parameters such as rotational wind speed,
translational wind speed, radius of maximum rotational wind speed, pressure
differential, and rate of pressure change.
Article 21. Assessment of flooding hazards
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2.
The potential for flooding in the area surrounding the site due to natural
causes, including storm surges, wind-generated waves, tsunamis, snowstorms, extreme
precipitation, and combined conditions arising from a common cause or having a
relatively high frequency of occurrence, shall be assessed.
3.
Appropriate meteorological, hydrological, and hydraulic models shall be
developed to determine flooding hazards, including secondary effects such as
debris and sediment. Relevant information from studies of historical and
prehistoric flooding shall be used to estimate the frequency and intensity of
flooding hazards.
4.
The potential instability of coastal areas or riverbeds due to erosion or
deposition shall be investigated.
5.
The potential occurrence of tsunamis or earthquakes in the area surrounding the
site that may adversely affect the safety of the nuclear power plant shall be
assessed.
6.
Hazards associated with tsunamis or earthquakes shall be derived from
historical data, available information on prehistoric flooding, physical
models, and analytical models. Such hazards shall include the potential for
water drawdown and water rise that may result in physical impacts on the site.
7.
Tsunami-related hazards relevant to the area surrounding the site shall be
assessed using nearshore bathymetry and coastal topography, taking into account
amplification caused by coastal configuration.
8.
Upstream water control structures, including dams, shall be assessed to
determine potential hazards associated with the failure of such upstream
structures, including cases involving combined conditions with flooding from
other causes.
9.
Where preliminary screening indicates that the nuclear power plant cannot
withstand the impacts of an accident involving an upstream water control
structure, the hazards to the nuclear power plant associated with such impacts
shall be assessed. The upstream structure shall be analyzed using a method
equivalent to that used to determine the hazards to the nuclear power plant to
demonstrate that the upstream structure can survive the relevant event.
10.
Flooding and related phenomena caused by water accumulation resulting from blockage
of rivers upstream or downstream, including blockage caused by landslides, or
from changes in land use, shall be considered.
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1.
The geotechnical characteristics and geological properties of subsurface
materials shall be investigated. The stratigraphic structure of soils and rocks
shall be characterized, taking into account variations and uncertainties in
subsurface materials.
2.
The static and dynamic geotechnical characteristics and geological properties
of subsurface materials at the site, including fill materials, shall be
established. Laboratory-based and in-situ methods, together with appropriate
sampling techniques and sufficient repetition of each test, shall be used to
characterize each parameter of the subsurface materials at the site.
3.
The stability and bearing capacity of foundation materials shall be assessed,
including consideration of the potential for excessive settlement under static
loading and earthquake conditions.
4.
The physical and geochemical properties of soil and groundwater shall be
investigated using appropriate methods.
Article 23. Assessment of geotechnical and geological hazards
1.
Geotechnical and geological hazards, including slope instability, uplift and
subsidence, soil liquefaction, and their impacts on the safety of the nuclear
power plant, shall be assessed.
2.
The site and the site vicinity shall be assessed to determine the potential for
slope instability, including landslides and rockfalls, caused by natural or
human-induced phenomena that may adversely affect the safety of the nuclear
power plant. The configuration of the site shall be taken into account when
assessing slope instability. Extreme meteorological conditions and rare
meteorological events shall also be taken into account when assessing slope
stability.
3.
The potential for slope instability due to seismic loading shall be assessed
using appropriate parameters characterizing earthquake hazards and the soil and
groundwater conditions at the site.
4.
The potential for surface uplift or subsidence that may adversely affect the
safety of the nuclear power plant throughout its operating lifecycle shall be
assessed using a detailed characterization of subsurface conditions obtained
through reliable investigation methods.
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6.
Accepted methods for in-situ and laboratory testing, in combination with
analytical methods, shall be used when assessing soil liquefaction.
Article 24. Assessment of other natural hazards
1.
Other natural phenomena characteristic of the area surrounding the site that
may adversely affect the safety of the nuclear power plant shall be
investigated.
2.
Other natural external hazards, including forest fires, droughts, hail, river
diversion, landslides, and biological hazards, shall be identified and assessed
as the basis for establishing the site design parameters for each hazard.
Section 3. REQUIREMENTS FOR SURVEY AND ASSESSMENT OF HUMAN-INDUCED
EXTERNAL HAZARDS
Article 25. Assessment of human-induced hazards
1.
Human-induced hazards at the site or in the area surrounding the site shall be
assessed.
2.
Human-induced hazards include:
a)
Hazards associated with road, inland waterway, maritime, or air transport,
including collisions and explosions;
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c)
Electromagnetic interference.
3.
Human activities such as resource extraction or other activities that
significantly alter land or water conditions, or reservoir-induced seismicity,
shall be considered.
4.
The likelihood of an aircraft crash at the site shall be assessed, taking into
account potential future changes in air traffic volume and aircraft
characteristics.
5.
Current or foreseeable activities in the area surrounding the site involving
the handling, processing, transport, or storage of chemicals capable of causing
explosions or forming flammable or explosive gas clouds shall be considered.
6.
Hazards associated with chemical explosions or other release events shall be
characterized in terms of heat, overpressure, and toxicity, where applicable,
taking into account the effects of distance and unfavorable combinations of
atmospheric conditions at the site. The potential impacts of such hazards on
personnel at the site shall be assessed.
Article 26. Dispersion of radioactive materials
1.
The extent of the dispersion through air and water of radioactive materials
released from the nuclear power plant under operating states and accident
conditions shall be assessed.
2.
Analysis of the atmospheric dispersion of radioactive materials shall take into
account relevant factors, including topography, land cover, and the
meteorological characteristics of the area surrounding the site, such as wind
speed and direction, air temperature, precipitation, humidity, atmospheric
stability parameters, and other meteorological parameters necessary for
modeling atmospheric radioactive dispersion. Long-term meteorological data for
locations near the site, where available, shall be collected, quality-assessed,
and used.
3.
A meteorological monitoring program shall be developed and implemented at the
site and in the area surrounding the site to assess atmospheric radioactive
dispersion. Monitoring equipment capable of recording meteorological parameters
at appropriate elevations, locations, and sampling intervals shall be used.
Data shall be collected for at least 12 consecutive months and used in the
analysis of atmospheric radioactive dispersion. In addition to monitoring data,
available meteorological data from other sources shall be collected.
Meteorological data shall be presented in terms of appropriate meteorological
parameters.
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5.
A surface-water study program shall be developed for the area surrounding the
site. The description of surface water shall include the principal physical and
chemical characteristics of the water, both natural and human-induced, the
principal water control structures, the locations of water intake structures,
information on seasonal variations, and water use in the area surrounding the
site.
6.
A hydrogeological study program shall be developed for the area surrounding the
site, including a description of the principal characteristics of aquifers and
their interaction with surface water, as well as data on groundwater use in the
area.
7.
The hydrogeological study program for the area surrounding the site shall include
studies of the transport and retention characteristics of radionuclides in
groundwater and studies of the associated exposure pathways.
8.
Hydrological and hydrogeological studies shall determine, to the extent
necessary, the dilution and dispersion characteristics of water, the potential
for reconcentration in sediments and organisms, the transport and retention
characteristics of radionuclides, the mechanisms for radionuclide transfer
within the hydrosphere, and the associated exposure pathways.
Article 27. Population distribution and public radiation exposure
1.
The current and projected population distribution in the area surrounding the
site, including permanent and transient populations throughout the operating
lifecycle of the nuclear power plant, shall be determined. Particular attention
shall be paid to vulnerable population groups and residential or institutional
facilities, including schools, hospitals, nursing homes, and prisons, when
assessing the potential impacts of radioactive releases and considering the
feasibility of implementing protective actions.
2.
The potential impacts of radioactive releases on the public under all operating
states and accident conditions shall be assessed once every 10 years and
updated whenever there is a significant change in population distribution. The
most recent population survey data for the area concerning permanent and
transient populations shall be used when assessing population distribution.
3.
Population distribution data shall be analyzed by direction and distance from
the site. Such information shall be used to assess the potential radiological
impacts of routine radioactive discharges and accidental radioactive releases,
including releases arising from severe accidents, using the site design parameters
and appropriate models.
Article 28. Land and water use in area surrounding site
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2.
Identification of land and water use characteristics shall include studies of
land resources, surface water, and groundwater that may be used by the
population or serve as habitats for organisms in the food chain.
Section 4. REQUIREMENTS FOR MONITORING AND PERIODIC ASSESSMENT
Article 29. Monitoring of external hazards and site conditions
1.
All natural and human-induced external hazards and site conditions relevant to
the licensing and safe operation of the nuclear power plant shall be monitored
throughout the operating lifecycle of the nuclear power plant.
2.
Monitoring of natural and human-induced external hazards and site conditions
shall commence no later than the start of construction and shall continue until
the operational termination of the nuclear power plant.
3.
A monitoring plan shall be developed, including the parameters to be monitored,
the types of data to be collected, methods of data collection, including the
locations and frequency of data collection, the required resolution and
accuracy of all measurements, data retention requirements, and requirements for
data processing and analysis. The monitoring plan shall be reviewed and
approved by the competent authority before implementation.
4.
Before the commencement of the trial operation of the nuclear power plant,
background radioactivity levels in air, soil, sediment, water, vegetation, and
foodstuffs in the area surrounding the site shall be measured to enable
identification of any increase in radioactivity resulting from the operation of
the nuclear power plant.
Article 30. Assessment of external hazards and site conditions
The
operating organization shall periodically assess all natural and human-induced
external hazards and site conditions as part of the periodic safety review
throughout the operating lifecycle of the nuclear power plant, in accordance
with Points d and dd Clause 1 Article 14 of Decree No. 316/2025/ND-CP, and
shall reassess the external hazards and site conditions, where necessary, on
the basis of the results of the periodic safety assessment of site-specific
hazards.
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SAFETY
REQUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS
Section 1. DESIGN PRINCIPLES
Article 31. Fundamental safety functions
1.
The fundamental safety functions of a nuclear power plant include: control of
the nuclear chain reaction; removal of heat from the reactor and from the spent
nuclear fuel storage area; confinement of radioactive materials; radiation
shielding; control of radioactive releases; and limitation of accidental
radioactive releases.
2.
The design of a nuclear power plant shall ensure the fundamental safety
functions prescribed in Clause 1 of this Article for all plant states.
3.
The design of a nuclear power plant shall provide means for monitoring the
plant state to ensure the performance of the safety functions prescribed in
Clause 1 of this Article.
Article 32. Radiation protection
1.
The design of a nuclear power plant shall ensure that radiation doses to plant
personnel and the public do not exceed the permissible limits and comply with
the ALARA principle under normal operation and throughout the design operating
lifetime of the plant; and are maintained below acceptable limits and in
accordance with the ALARA principle under accident conditions.
2.
The design shall satisfy the following requirements:
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b)
Plant states that may realistically occur shall result in no radiological
consequences or only insignificant radiological consequences.
3.
Acceptable limits for the purposes of radiation protection shall be established
in accordance with the categories of plant states and applicable laws.
Article 33. Design of nuclear power plants
1.
The design of a nuclear power plant and items important to safety shall
incorporate appropriate features to perform the safety functions with the
required reliability; ensure that the nuclear power plant can be safely
operated within the operational limits and conditions throughout its design
lifecycle; enable safe decommissioning; and minimize environmental impacts.
2.
Human capabilities and limitations and factors affecting human performance
shall be taken into account, and sufficient design information shall be
provided to ensure safe operation and maintenance and to enable upgrading and
modification where necessary.
3.
Relevant experience gained from the design, construction, and operation of
other nuclear power plants, as well as relevant research results, shall be
taken into account.
4.
The results of the deterministic safety analysis and probabilistic safety
analysis shall be used to ensure that accident prevention and mitigation of
accident consequences are adequately addressed in the design.
5.
The quantity and activity of generated radioactive waste and radioactive
releases shall be kept as low as reasonably achievable through appropriate
design measures, operating practices, and decommissioning arrangements.
Article 34. Application of defense-in-depth principle
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2.
The design shall ensure multiple levels of protection to prevent the
consequences of accidents that may cause harm to humans and the environment,
and shall ensure appropriate measures for protecting humans and the environment
and mitigating consequences where an accident cannot be prevented.
3.
The levels of defense in depth include:
a)
Level 1: prevention of deviations from normal operation and failures of items
important to safety;
b)
Level 2: detection and control of deviations from normal operation, and
prevention of anticipated operational occurrences from escalating into accident
conditions;
c)
Level 3: control of accidents within the design basis; prevention of damage to
the reactor core or radioactive releases requiring off-site protective
measures; and bringing the plant to a safe state;
d)
Level 4: prevention of accident progression resulting from the failure of Level
3 and mitigation of the consequences of severe accidents;
dd)
Level 5: mitigation of the consequences of releases of radioactive substances.
4.
The design shall ensure that the existence of multiple levels of protection is
not used as a basis for continued operation where one level of protection is
lost. All levels of defense in depth shall be maintained at all times, and any
relaxation shall be permitted only where clearly justified for a specific
operating mode.
5.
The design of a nuclear power plant shall satisfy the following requirements:
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b)
The plant shall be designed using a conservative approach and constructed to a
high standard of quality to minimize failures, prevent accidents to the
greatest extent possible, and ensure that minor deviations in nuclear power
plant parameters do not result in a cliff-edge effect;
c)
Inherent and technical features shall be provided so as to minimize or
eliminate the actuation of safety systems due to failures or deviations during
normal operation;
d)
Automatic safety systems shall be provided to control failures beyond the
capability of the control system and minimize the need for early intervention
by operating personnel;
dd)
Structures, systems and components, and procedures shall be provided to control
and limit the consequences of failures beyond the capability of the safety
systems;
e)
Multiple means shall be provided to ensure the performance of each fundamental
safety function, thereby ensuring the effectiveness of the barriers and
minimizing the consequences of failures or deviations during normal operation.
6.
The requirements for defense in depth shall be maintained by preventing, to the
greatest extent possible, the following:
a)
Impairment of the integrity of physical barriers;
b)
Failure of one or more barriers;
c)
Failure of one barrier resulting from the failure of another barrier;
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7.
The design shall ensure, to the greatest extent possible, that the first level
of protection, or at most the second level of protection, is capable of
preventing failures or deviations from normal operation that may occur
throughout the operating lifetime of the nuclear power plant from developing
into accident conditions.
8.
The levels of defense in depth shall be independent of one another to the
greatest extent possible so as to prevent the failure of one level from
reducing the effectiveness of other levels. Safety features for design
extension conditions, particularly features for mitigating the consequences of
fuel-melting accidents, shall be independent of the safety systems to the
greatest extent possible.
Article 35. Integration of safety, security, and safeguards
Safety
measures, nuclear security measures, and arrangements relating to nuclear
safeguards at a nuclear power plant shall be designed and implemented in an
integrated manner to ensure that such measures and arrangements do not
adversely affect one another.
Article 36. Application of proven designs
1.
Priority shall be given to the application of designs proven through
operational experience for items important to safety. Where this requirement
cannot be met, high-quality items incorporating qualified and tested
technologies shall be used.
2.
National standards, national technical regulations, international standards,
and foreign standards and technical regulations used as the design basis for
items important to safety shall be identified and assessed in terms of their
applicability, suitability, and adequacy; where necessary, they shall be
supplemented or adjusted to ensure that the design quality is commensurate with
the required safety function.
3.
Where an unproven design or technical feature is applied, or where a design
differs from a proven design, safety shall be demonstrated through a research
and testing program with specific acceptance criteria or through a review of
operating experience from similar applications. A new design or feature shall
be tested to the greatest extent possible before use and shall be monitored
during operation to verify compliance with the intended safety requirements.
Article 37. Safety assessment in nuclear power plant design
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2.
Safety assessments shall commence at an early stage of the design process and
shall be performed through an iterative approach between design activities and
confirmatory analyses, with their scope and level of detail increasing as the
design develops.
3.
Safety assessments shall be documented in a manner that facilitates independent
assessment where necessary.
Article 38. Requirements for construction assurance, waste management,
and decommissioning of nuclear power plants
1.
The design shall take into account the requirements necessary for future
construction, radioactive waste management, and decommissioning of the nuclear
power plant.
2.
The design shall take into account conditions during the construction stage to
ensure that construction can be carried out safely, meets quality criteria, and
conforms to the approved design.
3.
The design shall include measures necessary to minimize the quantity and
activity of waste generated during operation, maintenance, and decommissioning
of the plant, as well as measures to ensure the safe treatment, storage, and
disposal of radioactive waste.
4.
The design shall take into account the technical requirements and supporting
structures, systems, and components necessary for the safe and effective
decommissioning of the nuclear power plant, while minimizing environmental
impacts to the greatest extent possible.
Section 2. GENERAL REQUIREMENTS
Article 39. Classification of nuclear power plant states
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a)
Normal operation;
b)
Anticipated operational occurrences;
c)
Design basis accidents;
d)
Design extension conditions, including severe accidents.
2.
The design shall establish specific criteria for each plant state, ensuring
that frequently occurring states result in no radiological consequences or only
minor radiological consequences, and that states capable of resulting in severe
consequences have a very low frequency of occurrence.
Article 40. Design basis and design limits for items important to safety
1.
The design shall clearly establish the design basis, including the required
capability, reliability, and functions of items important to safety for each operating
state, accident condition, and internal and external hazard, so as to ensure
compliance with the acceptance criteria throughout the plant lifecycle.
2.
The design shall specify the design limits for each item important to safety,
in accordance with the applicable standards and technical regulations, so as to
ensure that the required safety functions are maintained under all conditions
identified in Clause 1 of this Article.
3.
The design basis and design limits shall be demonstrated, systematically
verified, documented, and supported by sufficient information to ensure safe
operation.
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1.
A systematic approach shall be applied to identify and take into account in the
design all postulated initiating events, so that every event expected to be
capable of causing severe consequences and every event expected to have a
significant frequency of occurrence is adequately considered.
2.
Postulated initiating events shall be identified on the basis of technical
assessment in combination with deterministic safety analysis and probabilistic
safety analysis. The application of these methods shall be justified as
providing adequate consideration of all potential initiating events.
3.
Postulated initiating events shall include all anticipated failures of
structures, systems, and components, operating errors, and failures caused by
internal and external hazards, under all damage states that may arise from
internal and external hazards at high-power states, low-power states, and reactor
shutdown states.
4.
Postulated initiating events shall be analyzed to determine the preventive and
protective measures necessary to ensure the performance of the required safety
functions.
5.
The design shall ensure that the plant responds to postulated initiating events
in the following order of priority with respect to the level of safety:
a)
Inherent plant characteristics automatically cause the event to progress in a
safe direction or ensure that the event has no significant effect on safety;
b)
Passive safety features or systems in continuous operation automatically
control the event and bring the plant to a safe state;
c)
Dedicated safety systems are actuated to bring the plant to a safe state;
d)
Operating procedures are implemented by personnel to bring the plant to a safe
state.
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7.
Technical justification shall be provided at the design stage for excluding
initiating events that are not included in the list of postulated initiating
events.
8.
Where a rapid and reliable response to a postulated initiating event is required,
the design shall provide for the automatic actuation of safety systems to
prevent progression to more severe plant conditions.
9.
Where a rapid response to a postulated initiating event is not required, manual
initiation of systems or other actions by operating personnel may be permitted,
subject to the following conditions:
a)
The time interval between detection of the event and the required action is
sufficiently long;
b)
Appropriate administrative, operating, and emergency response procedures are
available;
c)
The possibility of personnel aggravating the accident through incorrect action
or misdiagnosis has been assessed and excluded.
10.
The design shall provide sufficient instrumentation, condition monitoring, and
controls to support operating personnel in assessing the plant state following
an initiating event and in taking actions to bring the plant to a stable,
long-term shutdown state.
11.
The design shall identify the equipment and procedures necessary to ensure the
capability to control the plant and mitigate consequences in the event of loss
of control following a postulated initiating event.
12.
Manual equipment for accident response and recovery shall be located at
appropriate locations, maintained in a state of readiness, and capable of being
safely accessed under the environmental conditions anticipated during an
accident.
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1.
The impacts of all internal and external hazards, including human-induced
events that may directly or indirectly affect plant safety, shall be identified
and assessed. The results of the hazard assessment shall serve as the basis for
the design of the site layout, building spaces, and equipment systems, the
identification of postulated initiating events, and the determination of load
combinations for the design of items important to safety.
2.
The impacts of all internal hazards shall be identified and assessed in
accordance with their significance to safety, so as to ensure the capability to
withstand the effects of such hazards and to prevent common-cause failures
caused by them.
3.
For sites with multiple units, the design shall fully consider the possibility
that specific hazards may simultaneously affect multiple units or all units at
the same site.
4.
The design shall fully consider internal hazards or hazards originating from
other facilities at the same site and shall incorporate appropriate preventive
and mitigating features to ensure that safety is not adversely affected.
5.
The design shall fully consider natural or human-induced external events
identified during the site assessment. The causes and probabilities of
occurrence shall be considered when postulating potential hazards. The design
shall ensure that short-term plant safety does not depend on the availability
of off-site resources and shall fully consider site-specific conditions in
determining the maximum delay before off-site resources must become available.
6.
The design shall incorporate technical features to minimize adverse interactions
between buildings containing items important to safety and other plant
buildings during external events.
7.
The design shall provide an adequate safety margin to protect items important
to safety against cliff-edge effects arising from the external hazards
identified in the design.
8.
The design shall provide an adequate safety margin to protect items serving as
the final barrier necessary to prevent early radioactive releases or large
radioactive releases into the environment, even where the severity of a natural
hazard exceeds the design level established through the site assessment.
Article 43. Technical design rules
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2.
The design shall apply methods to ensure design robustness and shall be
consistent with proven designs so as to ensure that the fundamental safety
functions are achieved for all operating states and accident conditions.
Article 44. Design basis accidents
1.
Design basis accident conditions shall be identified from postulated initiating
events to establish boundary conditions ensuring that the nuclear power plant
is capable of withstanding such conditions without exceeding the prescribed
radiation protection limits.
2.
Design basis accidents shall be used to establish the design basis for safety
systems and other items important to safety for controlling design basis
accident conditions, with the objective of bringing the plant back to a safe
state and mitigating the consequences of any accident.
3.
The design shall ensure that, under design basis accident conditions, the
principal plant parameters do not exceed the specified design limits, with the
objective of managing design basis accidents such that they result in no
radiological consequences or only insignificant radiological consequences at or
beyond the plant site, and do not require the implementation of off-site
emergency response measures.
4.
Design basis accidents shall be analyzed using a conservative approach.
Article 45. Design extension conditions
1.
A set of design extension conditions shall be developed on the basis of
technical assessment, deterministic safety analysis, and probabilistic safety
analysis to enhance the capability of the plant to withstand accidents more
severe than design basis accidents, or accidents involving additional failures
occurring simultaneously or sequentially. Design extension conditions shall be
used to identify accident scenarios to be considered in the design and to
establish plans for preventing or mitigating accident consequences.
2.
The analysis of design extension conditions shall aim to ensure that the plant
design is capable of preventing accident conditions not considered in design
basis accidents, or mitigating their consequences in accordance with the ALARA
principle. The results of the analysis may lead to requirements for additional
safety features for design extension conditions or for extending the
capabilities of safety systems to prevent or mitigate the consequences of severe
accidents or to maintain the radioactive confinement function.
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4.
Design extension conditions shall be used to determine the design
specifications for safety features and items important to safety necessary to
prevent, control, and mitigate accident consequences.
5.
The analysis of design extension conditions shall identify the features
designed to prevent or mitigate events under design extension conditions. Such
features shall:
a)
Be independent, to the extent practicable, of the features used for accidents
having a higher frequency of occurrence;
b)
Be capable of functioning under environmental conditions associated with design
extension conditions, including severe accident conditions;
c)
Have reliability commensurate with their functions.
6.
The reactor bunker and its safety features shall be capable of withstanding
extreme scenarios, including reactor core melting. Such scenarios shall be
selected using engineering assessment and probabilistic safety analysis.
7.
The design shall ensure that the possibility of conditions that could lead to
an early radioactive release or a large radioactive release is practically
eliminated.
8.
The design shall ensure that, for design extension conditions, the duration and
extent of protective measures for the population are minimized, while allowing
sufficient time for such measures to be implemented.
9.
Where technical assessment, deterministic safety analysis, and probabilistic
safety analysis indicate that combinations of events may lead to anticipated
operational occurrences or accident conditions, such combinations shall be
classified as design basis accidents or design extension conditions, depending
on their likelihood of occurrence. Secondary events arising from an initial
event shall be considered part of the original postulated initiating event.
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1.
Interactions between safety systems, or between redundant channels of a system,
shall be prevented by measures such as physical separation, electrical
isolation, functional independence, and independence of signal transmission.
2.
Equipment of a safety system, including cables and cable conduits, shall be
marked to facilitate identification of each redundant channel of the safety
system.
Article 47. Safety classification
1.
All items important to safety shall be identified and classified on the basis
of their functions and significance to safety.
2.
The classification of items important to safety according to their significance
to safety shall be based primarily on a deterministic approach, supplemented by
a probabilistic approach where appropriate, with full consideration given to
the following factors:
a)
The safety function to be performed by the item;
b)
The consequences of failure to perform the safety function;
c)
The frequency with which the item is required to perform the safety function;
d)
The time following a postulated initiating event at which the item will be
required to perform the safety function, or the period for which the item is
required to maintain that function.
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4.
Equipment performing multiple functions shall be classified in the safety class
corresponding to the most important function performed by such equipment.
Article 48. Reliability of items important to safety
The
reliability of items important to safety shall be commensurate with their
significance to safety, as demonstrated through compliance with the following
requirements:
1.
Items important to safety shall be assessed and quality-assured throughout
procurement, installation, acceptance testing, operation, and maintenance so as
to be capable of withstanding design basis accidents.
2.
In selecting equipment, inadvertent actions and the possibility of failure
shall be taken into account, with preference given to equipment that can be
readily repaired or replaced in the event of failure.
Article 49. Common-cause failures, single failure criterion, and
fail-safe design
1.
Equipment design shall give full consideration to the potential for common-cause
failures of items important to safety to determine the manner in which the
concepts of diversity, redundancy, physical separation, and functional
independence are to be applied to achieve the required reliability.
2.
The single failure criterion shall be applied to each safety group incorporated
into the plant design.
3.
Unauthorized or unintended actions resulting from signal failures or equipment
failures shall be considered a form of failure when applying the single failure
criterion to a safety group or safety system.
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5.
Systems and components important to safety shall be designed in accordance with
the fail-safe principle, where appropriate.
6.
Systems and components important to safety shall be appropriately designed to
achieve a safe state in the event of failure, ensuring that their failure, or
the failure of a supporting feature, does not result in the loss of capability
to perform the intended safety function.
Article 50. Support service systems
1.
Support service systems forming part of a system important to safety shall be
assigned a safety classification commensurate with the significance to safety
of the supported system.
2.
The reliability, degree of redundancy, diversity, and independence of support
service systems, as well as the provision of features enabling their isolation
and testing of their functional capability, shall be commensurate with the
significance to safety of the supported system.
3.
A failure of a support service system shall not be capable of simultaneously
affecting redundant channels or components of a safety system, or a system
performing different safety functions, in a manner that impairs the capability
of those systems to perform their safety functions.
Article 51. Operational limits and conditions for safe operation
The
design of a nuclear power plant shall establish a set of operational limits and
conditions at the design stage to ensure the safe operation of the nuclear power
plant, including:
1.
Safety limits.
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3.
Operating limits and conditions for normal operation.
4.
Set points for the control system and procedural limits for process variables
and other important parameters.
5.
Surveillance, maintenance, testing, and inspection requirements to ensure that
structures, systems, and components perform their functions as designed, in
accordance with optimization requirements and the ALARA principle.
6.
Establishment of operating configurations, including operating limits where a
safety system or safety-related system is unavailable.
7.
Actions to be taken, and the time allowed for such actions, in the event of
deviations from the operational limits and conditions.
Article 52. Calibration, testing, maintenance, repair, replacement,
inspection, and monitoring of items important to safety
The
design of items important to safety shall ensure the following requirements:
1.
Facilitate calibration, testing, maintenance, repair, replacement, inspection,
and monitoring of their capability to perform their functions and maintain
their integrity under all conditions identified in the design basis.
2.
Ensure that calibration, testing, maintenance, repair, replacement, inspection,
and monitoring activities do not result in excessive radiation exposure to the
persons performing such activities.
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4.
Where it is not possible to design items important to safety to satisfy the
requirements for direct testing, inspection, or monitoring to the necessary
extent, reliable technical justification shall be provided using the following
approaches:
a)
Indirect testing, inspection, or monitoring methods shall be available through
reference items, using validated computational methods with predictive
capability to substitute for such items;
b)
Adequate safety margins shall be provided, or appropriate preventive measures
shall be implemented to compensate for potential failures.
Article 53. Quality assurance for items important to safety
1.
A quality assurance program shall be implemented for items important to safety
to demonstrate that such items are capable of performing the required functions
under the current environmental conditions and under changes in environmental
conditions anticipated in the design basis throughout the design lifecycle.
2.
The quality assurance program for items important to safety shall include
consideration of ageing effects caused by environmental factors. Where items
important to safety are affected by natural external hazards, consideration
shall be given to a qualification program for such items under similar
conditions that have occurred.
3.
The quality assurance program for items important to safety shall take into
account all adverse environmental conditions that may arise during the
operation of the nuclear power plant.
Article 54. Ageing management
The
design of the nuclear power plant shall ensure the following requirements:
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2.
The effects of ageing and wear shall be fully considered under all operating states
of equipment, including testing, maintenance, maintenance outages, and plant
states during and after a postulated initiating event.
3.
Measures for monitoring, testing, sampling, and inspection shall be provided to
assess ageing mechanisms identified at the design stage and to identify adverse
changes in the nuclear power plant or degradation occurring during plant
operation.
Article 55. Optimization of operating personnel performance
1.
Human factors, including the human-machine interface, shall be systematically
assessed, incorporated into the design from an early stage of the design
process, and maintained throughout the design process.
2.
The design of the nuclear power plant shall determine the minimum number of
operating personnel necessary to simultaneously perform actions required to
bring the plant to a safe state in the event of an operating state or accident
condition.
3.
The design shall take into account the experience of operating personnel at
similar nuclear power plants and shall support operating personnel in assessing
and responding to situations during the operation of the nuclear power plant
and maintenance of equipment.
4.
The design shall optimize the performance of the responsibilities of operating
personnel and minimize safety impacts resulting from operating errors. The
design process shall consider the overall plant layout, equipment, and
operating procedures to facilitate interaction between operating personnel and
the plant under all states.
5.
The human-machine interface design shall provide operating personnel with
sufficient and manageable information, appropriate to the time required for
decision-making and implementation of actions. Information necessary for
decision-making shall be presented simply and clearly, including information
for:
a)
Assessing the condition of the plant;
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c)
Confirming that automatic safety actions are actuated when necessary and that
the relevant systems operate correctly as designed;
d)
Determining the need for, and timing of, manual actuation of identified safety
actions.
6.
The design shall enhance the likelihood of successful actions by operating
personnel, taking into account the time required to perform the actions,
anticipated conditions, and psychological factors affecting operating personnel.
7.
The design shall minimize the need for operating personnel to take actions
within a short period of time and shall justify that operating personnel have
sufficient time to make decisions and perform the required actions.
8.
The design shall ensure that, following an event affecting the plant,
environmental conditions in the main control room, supplementary control room,
and access routes to these areas do not adversely affect the safety of
operating personnel.
9.
The design of the workplace and working environment of operating personnel
shall comply with ergonomic principles.
10.
Verification and validation of human-factor-related features shall be performed
to confirm that the required actions of operating personnel have been
identified and can be performed correctly.
Article 56. Safety systems and safety features for design extension
conditions at multi-unit nuclear power plants
1.
Each unit of a nuclear power plant shall have its own safety systems and safety
features for design extension conditions.
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Article 57. Requirements for radioactive material confinement systems
Systems
in a nuclear power plant that may contain radioactive materials, including
fissile materials, shall be designed to:
1.
Prevent accidents that may result in uncontrolled radioactive releases into the
environment.
2.
Prevent unintended criticality and overheating.
3.
Ensure that radioactive releases are maintained below permissible discharge
limits during normal operation and below acceptable limits under accident
conditions, while complying with the ALARA principle.
4.
Mitigate radiological consequences in the event of an accident.
Article 58. Escape routes from plant
1.
The nuclear power plant shall be provided with sufficient escape routes that
are clearly indicated and equipped with emergency lighting, ventilation
systems, and other necessary means to ensure safety when the escape routes are
used.
2.
Escape routes from the nuclear power plant shall comply with requirements
applicable to radiation areas, fire and explosion protection, industrial
safety, and nuclear security of the nuclear power plant.
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Article 59. Requirements for communication systems
1.
Diverse and effective means of communication shall be provided throughout the
plant, enabling both internal and external communication and remaining usable
under all circumstances.
2.
Appropriate alarm systems shall be provided to issue warnings and instructions
under operating states and accident conditions.
Article 60. Control of access to plant and prevention of unauthorized
access to or interference with items important to safety
1.
Appropriate measures shall be provided to isolate the nuclear power plant from
the surrounding area, with different measures available to control access to
the plant.
2.
The measures prescribed in Clause 1 of this Article shall take into account the
need for access to the nuclear power plant in the event of an accident and for
the implementation of emergency response measures.
3.
The measures prescribed in Clause 1 of this Article shall take into account the
capability to control movement and prevent unauthorized access to or interference
with plant equipment, particularly items important to safety, including
computer hardware and software.
Article 61. Prevention of harmful interactions between systems important
to safety
1.
Hazards that may arise from interactions between systems important to safety
during simultaneous operation shall be assessed and prevented.
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3.
Where two systems containing fluids important to safety are interconnected and
operate at different pressures, both systems shall be designed to withstand the
higher pressure, or provisions shall be made to prevent the design pressure of
the system operating at the lower pressure from being exceeded.
Article 62. Interaction between power grid and plant
The
capability of items important to safety in the nuclear power plant to perform
their functions shall not be adversely affected by fluctuations in the power
grid, including anticipated variations in the voltage and frequency of the
power supply from the grid.
Article 63. Safety analysis in plant design
1.
Safety analysis shall be performed for the design of the nuclear power plant,
applying both deterministic safety analysis and probabilistic safety analysis
to assess and consider hazards and factors that may affect safety under
different plant states.
2.
On the basis of the safety analysis, the design basis of items important to
safety and their relationship with initiating events and event sequences shall
be confirmed. It shall be demonstrated that the nuclear power plant, as
designed, is capable of complying with the permissible limits for radioactive
releases and radiation dose limits under all operating states, as well as
satisfying the acceptable limits applicable to accident conditions.
3.
The safety analysis shall provide evidence that the defense-in-depth principle
has been applied in the design.
4.
The safety analysis shall ensure that uncertainties are adequately taken into
account in the design and that sufficient safety margins are provided to avoid
cliff-edge effects and prevent early or large radioactive releases.
5.
The suitability of the assumptions, methods, and degree of conservatism used in
the plant design shall be updated and verified for the current design or the
as-built design after construction.
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a)
Establishment and confirmation of the design basis for all items important to
safety;
b)
Identification of the characteristics of postulated initiating events
appropriate to the site and plant design;
c)
Analysis and assessment of event sequences arising from postulated initiating
events to confirm qualification requirements;
d)
Comparison of analysis results with acceptance criteria, design limits,
radiation dose limits, and acceptable limits for radiation protection;
dd)
Demonstration that anticipated operational occurrences and design basis
accidents can be managed through automatic safety features and prescribed
actions by operating personnel;
e)
Demonstration that design extension conditions can be managed through automatic
actuation of safety systems, use of safety features, and anticipated actions by
operating personnel.
7.
The results of probabilistic safety analysis shall be fully taken into account
in the design for all operating modes and plant states, including reactor
shutdown states, in order to:
a)
Ensure that the overall plant risk is not dominated by any single design
feature or initiating event, and that the levels of defense in depth are
independent of one another to the extent practicable;
b)
Demonstrate that cliff-edge effects are prevented;
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Section 3. DESIGN OF NUCLEAR POWER PLANT SYSTEMS
Subsection 1. DESIGN OF REACTOR CORE AND RELATED FEATURES
Article 64. Requirements for fuel rods and fuel assemblies
1.
Fuel rods and fuel assemblies shall maintain their integrity under all
conditions and shall be capable of withstanding radiation and conditions within
the reactor core, including degradation after a period of use.
2.
Factors to be taken into account when considering the condition of fuel rods
and fuel assemblies after a period of use shall include:
a)
Expansion and deformation;
b)
External pressure of the coolant;
c)
Internal pressure caused by fission products and helium accumulation;
d)
Effects of irradiation;
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e)
Chemical effects;
g)
Static and dynamic loads; flow-induced vibration and mechanical vibration;
h)
Changes in heat transfer capability resulting from deformation or chemical
effects.
3.
The design shall take into account margins for uncertainties in data,
calculations, and manufacturing processes.
4.
The fuel design limits shall include limits on the permissible leakage level of
fission products under anticipated operational occurrences, below which the
fuel may continue to be used safely.
5.
Fuel rods and fuel assemblies shall be capable of withstanding loads and stresses
arising during handling, installation, removal, transport, and storage as
prescribed.
Article 65. Requirements for reactor core structures to ensure
coolability and insertion of control rods into core
The
geometrical design of fuel rods, fuel assemblies, and supporting structures
shall ensure that cooling capability is maintained and that insertion of
control rods into the reactor core is not impeded during normal operation or in
the event of an accident, except for a severe accident.
Article 66. Reactor core control
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The
need to use control systems to maintain the shape, level, and stability of the
neutron flux within the specified design limits under all operating states
shall be minimized.
2.
Sufficient means shall be provided to monitor the neutron flux distribution in
the reactor core, ensuring that the neutron flux at any location within the
core does not exceed the design limits.
3.
The design of reactivity control devices shall take into account degradation of
the devices due to irradiation effects, burnup, changes in physical properties,
and gas generation.
4.
The maximum positive reactivity and the rate of reactivity increase under
operating states and accident conditions shall be limited or compensated so as
to ensure that no significant fuel damage occurs, prevent failure of the
reactor coolant pressure boundary, maintain cooling capability, and prevent
significant damage to the reactor core.
Article 67. Reactor shutdown
1.
Means shall be provided to ensure the capability to shut down the reactor under
all conditions, including when the reactor has the highest positive reactivity.
2.
The effectiveness, speed, and shutdown margin shall be sufficient to ensure
that the fuel design limits are not exceeded.
3.
In assessing the effectiveness of reactor shutdown means, all failures in the
nuclear power plant that could partially disable the reactor shutdown means or
result in common-cause failures shall be taken into account.
4.
Reactor shutdown means shall satisfy the following requirements:
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b)
The systems shall prevent reactivity increases considered in the design from
resulting in unintended criticality during fuel loading, during reactor
shutdown, or while the reactor is in a shutdown state.
5.
Instrumentation and inspection provisions shall be provided to ensure that the
reactor shutdown means are maintained in the specified condition for each plant
state.
Subsection 2. REACTOR COOLANT SYSTEM
Article 68. Design of reactor coolant system
1.
Components of the reactor coolant system shall be designed and constructed to
minimize the risk of failure resulting from inadequate material quality,
inappropriate design standards, insufficient inspection or testing capability,
or poor manufacturing quality.
2.
Piping connections at the reactor coolant pressure boundary shall be
appropriately designed and fabricated to prevent coolant leakage through joints
and prevent the release of radioactive coolant.
3.
The design of the reactor coolant pressure boundary shall ensure that defects
are unlikely to develop; where cracks occur, they shall propagate slowly, shall
not result in sudden failure, and shall allow timely detection.
4.
Components of the reactor coolant pressure boundary shall not become embrittled
as a result of material degradation.
5.
Failure of a component within the reactor coolant pressure boundary shall not
result in failure of other items important to safety under all operating states
and under design basis accident conditions, taking into account their
degradation.
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The
design shall ensure that the operation of pressure relief devices provides
protection against overpressure at all locations of the reactor coolant
pressure boundary and does not result in direct radioactive releases from the
nuclear power plant into the environment.
Article 70. Control of reactor coolant
1.
The design of the nuclear power plant shall provide measures to control the
quantity, temperature, and pressure of the reactor coolant so as to ensure that
the design limits are not exceeded under all operating states of the nuclear
power plant, taking into account changes in coolant volume and coolant leakage.
2.
The design of the nuclear power plant shall provide a system for removing
radioactive substances (including activated corrosion products and fission
products released from the fuel) and non-radioactive substances from the
reactor coolant.
3.
The capability of the system referred to in Clause 2 of this Article shall be
based on the design limits for fuel leakage, with sufficient margin to ensure
that radioactivity in the primary circuit remains low, radioactive releases
remain below the permissible limits, and the ALARA principle is complied with.
Article 71. Removal of residual heat from reactor core
The
design of the nuclear power plant shall provide means for removing residual
heat from the reactor core when the nuclear power plant is in a shutdown state,
ensuring that the design limits for the fuel, the reactor coolant pressure
boundary, and structures important to safety are not exceeded.
Article 72. Emergency cooling of reactor core
1.
Means shall be provided for cooling the reactor core, restoring and maintaining
fuel coolability under accident conditions, including where the integrity of
the pressure boundary of the primary core cooling system cannot be maintained.
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a)
Limiting parameters relating to the integrity of the fuel cladding and the fuel
shall not be exceeded;
b)
Potential chemical reactions shall be maintained at an acceptable level;
c)
The means for cooling the reactor core shall be capable of compensating for
potential changes in the fuel and the internal geometrical configuration of the
core;
d)
Cooling of the reactor core shall be ensured for the required period of time.
3.
Appropriate design features, redundancy, and diversity shall be provided to
satisfy the requirements prescribed in Clause 2 of this Article with
reliability commensurate with each postulated initiating event.
Article 73. Heat transfer to ultimate heat sink
1.
The capability to transfer heat to the ultimate heat sink shall be ensured for
all plant states.
2.
Heat transfer systems shall have reliability commensurate with the operating
state in which the heat transfer function is required to be performed. Where
necessary, an alternative ultimate heat sink or an alternative means of access
to the ultimate heat sink may be used.
3.
The heat transfer function shall be ensured for levels of natural hazards more
severe than those considered in the design, on the basis of the results of the
site hazard assessment.
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Article 74. Functions of reactor bunker system
The
design of the reactor bunker system shall perform the following safety
functions:
1.
Prevent releases of radioactive substances into the environment under operating
states and accident conditions.
2.
Protect the reactor against natural and human-induced external hazards.
3.
Provide radiation shielding under operating states and accident conditions.
Article 75. Control of radioactive releases from reactor bunker
1.
The design of the reactor bunker shall ensure compliance with the ALARA
principle and minimize radioactive releases from the nuclear power plant into
the environment, ensuring that such releases remain below the permissible
discharge limits under operating states and below the acceptable limits under
accident conditions.
2.
Structures, systems, and components affecting the leak-tightness of the reactor
bunker system shall be designed, fabricated, and installed so as to allow
leakage testing at the design pressure throughout the operating lifetime of the
nuclear power plant.
3.
The design of reactor bunker penetrations shall satisfy the following
requirements:
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b)
The penetrations shall be capable of withstanding forces caused by
displacement, pipe whipping, or other accidents involving missiles, as well as
internal and external hazards.
Article 76. Reactor bunker isolation
1.
Piping penetrating the reactor bunker that forms part of the reactor coolant
pressure boundary or is directly connected to the atmosphere inside the reactor
bunker shall satisfy the following requirements:
a)
It shall be capable of automatic isolation with high reliability in the event
of an accident;
b)
The isolation capability prescribed in Point a of this Clause shall be provided
by at least two reactor bunker isolation valves or check valves installed in
series, normally with one valve inside and one valve outside the reactor
bunker, together with a leakage detection system. The isolation valves or check
valves shall be located as close as practicable to the reactor bunker; each
valve shall be capable of independent and reliable actuation and shall be
periodically tested;
c)
The requirements prescribed in Points a and b of this Clause shall not apply to
piping containing instrumentation or where application of such requirements
would reduce the reliability of a safety system penetrating the reactor bunker.
2.
Piping penetrating the reactor bunker that does not form part of the reactor
coolant pressure boundary and is not directly connected to the atmosphere
inside the reactor bunker shall be required to have at least one isolation
valve located outside the reactor bunker as close as practicable to the reactor
bunker.
Article 77. Access to reactor bunker
1.
Access doors to the reactor bunker for operating personnel shall be airtight.
Such doors shall operate through an interlocking mechanism to ensure that at
least one door remains closed under all circumstances. Means shall be provided
for monitoring access doors and access passages. The design shall include
requirements relating to personnel safety.
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Article 78. Control of reactor bunker conditions
1.
Measures shall be provided to control pressure and temperature within the
reactor bunker, as well as the accumulation of fission products or other
gaseous, liquid, and solid substances generated within the reactor bunker that
may affect the operation of systems important to safety.
2.
Adequate flow paths shall be provided between the different compartments within
the reactor bunker. The cross-sectional areas of openings between compartments
shall be calculated so that pressure differentials during pressure equalization
following an accident do not cause serious damage to pressure-retaining
structures or to systems important for mitigating accident consequences.
3.
The capability to remove heat from the reactor bunker shall be ensured to
reduce pressure and temperature and maintain them at acceptable levels
following an accident involving the release of high-energy fluids. Systems
performing the heat removal function from the reactor bunker shall have high
reliability and redundancy.
4.
Design measures shall be provided to prevent loss of the structural integrity
of the reactor bunker under all plant states, ensuring that no early
radioactive release or large radioactive release occurs.
5.
The design shall include features enabling the safe use of non-permanently
installed equipment to restore the capability to remove heat from the reactor
bunker.
6.
A system shall be provided to control fission products, hydrogen, oxygen, and
other substances that may be released into the reactor bunker to reduce the
quantity of fission products that may be released into the environment under
accident conditions.
Subsection 4. INSTRUMENTATION AND CONTROL SYSTEMS
Article 79. Instrumentation system
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2.
Instrumentation shall provide sufficient information to monitor the condition
of the nuclear power plant, the progression of an accident, and the location
and quantity of radioactive releases, and to support post-accident analysis.
Article 80. Control system
The
control system shall have sufficient reliability and capability to maintain the
relevant process parameters within the specified operating ranges.
Article 81. Protection system
1.
A protection system shall be provided that is capable of detecting unsafe plant
conditions and automatically actuating the necessary systems to achieve and
maintain the safe state of the plant.
2.
The protection system shall be designed to:
a)
Be capable of overriding unsafe actions of the control system;
b)
Incorporate fail-safe features so as to bring the nuclear power plant to a safe
state in the event of a failure of the protection system;
c)
Prevent actions by operating personnel from compromising the effectiveness of
the protection system, while not preventing appropriate actions by operating
personnel in the event of an accident;
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dd)
Provide operating personnel with the information necessary to monitor the
effects of automatic actions.
Article 82. Reliability and testability of instrumentation and control
systems
1.
Instrumentation and control systems for items important to safety shall be
designed with high reliability and with capability for periodic testing
commensurate with the safety functions of such items.
2.
Instrumentation and control systems shall be designed to facilitate testing,
provide automatic indication of system faults, and automatically remedy faults;
they shall incorporate diversity in functions and operating principles so as to
preserve the safety functions under all circumstances.
3.
Safety systems shall be designed to permit periodic testing of system
functions, including while the nuclear power plant is in operation, including
the capability to independently test individual channels to detect failures and
preserve redundancy, and to perform functional testing of sensors, input
signals, actuation mechanisms, and displays.
4.
Where a safety system or part thereof is taken out of service for testing, a
clear indication that it has been taken out of service for such testing shall
be provided.
Article 83. Use of computer-based equipment in systems important to
safety
1.
Where a system important to safety at a nuclear power plant relies on
computer-based equipment, provisions shall be established for the development
and testing of computer hardware and software throughout the lifecycle of the
system, particularly with respect to the software development process. A
quality management system shall be established for the entire development
process.
2.
Computer-based equipment used in safety systems and safety-related systems
shall satisfy the following requirements:
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b)
The design process for such equipment shall be systematically documented
throughout the entire design process, including control, testing, and trial
operation following design changes; the documentation shall be regularly
reviewed;
c)
The equipment shall be assessed by experts independent of the design team and
the supplier to ensure high reliability;
d)
The principle of diversity shall be applied to the design of equipment
important to safety where its high reliability has not been clearly justified;
dd)
Common-cause failures of equipment originating from computer software shall be
taken into account;
e)
The equipment shall be protected against inadvertent disruption or deliberate
interference with system operation.
Article 84. Separation of protection system and control system
1.
The protection system and the control system shall be designed to be
functionally independent, and interactions between them shall be prevented by
separation measures.
2.
Where the protection system and the control system share signals, such signals
shall be classified as part of the protection system. The separation of the two
systems in such case shall be clearly justified.
Article 85. Main control room
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2.
Appropriate barriers and separation measures shall be provided between the main
control room and the external environment. Sufficient information shall be
provided to protect personnel working in the main control room against hazards
arising during an accident, such as high radiation levels, releases of
radioactive substances, fire, explosions, or the dispersion of toxic gases.
3.
The design shall clearly identify internal and external events that could
threaten the continued operability of the main control room and shall provide
reasonable measures to mitigate the consequences of such events.
4.
The design of the main control room shall provide an appropriate safety margin
against natural hazards having a severity exceeding the levels considered in
the design, on the basis of the results of the site hazard assessment.
Article 86. Supplementary control room
1.
A supplementary control room shall be provided with instrumentation and controls
that are physically separated, electrically isolated, and functionally
independent from the main control room prescribed in Article 85 of this
Circular.
2.
The supplementary control room shall be capable of maintaining the reactor in a
safe shutdown state, removing residual heat, and monitoring changes in
parameters in the event that the relevant functions cannot be performed from
the main control room.
3.
The requirements for personnel protection prescribed in Clause 2 Article 85 of
this Circular shall apply to the supplementary control room.
Article 87. On-site emergency response center
1.
The nuclear power plant shall have an on-site emergency response center
separate from the main control room and the supplementary control room.
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3.
The center shall be equipped with means of communication with the main control
room, the supplementary control room, other important locations within the
nuclear power plant, and emergency response organizations.
4.
Measures shall be implemented to protect personnel working at the center for
extended periods against hazards arising during an accident.
5.
The center shall be provided with the systems, equipment, and conditions
necessary to enable emergency response personnel to work for extended periods.
Subsection 5. EMERGENCY POWER SUPPLY SYSTEM
Article 88. Emergency power supply system
1.
The design shall include:
a)
An emergency electrical cabling system having sufficient capability to provide
the electrical power required during anticipated operational occurrences and
design basis accidents in the event of loss of off-site power;
b)
An additional backup power source to provide sufficient electrical power
required under design extension conditions.
2.
The design specifications for the emergency power source and the backup power
source shall include requirements concerning power supply capability,
availability, required duration, capacity, and continuity of power supply.
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4.
The additional backup power source shall be capable of providing sufficient
electrical power to maintain the integrity of the reactor core cooling system
and prevent severe damage to the reactor core and spent nuclear fuel in the
event that loss of off-site power occurs simultaneously with failure of the
emergency power source.
5.
It shall be ensured that equipment necessary to mitigate the consequences of
reactor core melting can be supplied with electrical power from any remaining
available power source.
6.
The additional backup power source shall be independent of and physically
separated from the emergency power source. The connection time of the
alternative power source shall be compatible with the battery endurance time.
7.
Continuity of the electrical power supply required for monitoring principal
plant parameters and performing short-term actions necessary for safety shall
be maintained in the event of loss of alternating-current power sources.
8.
The design of diesel engines and generators providing emergency power to items
important to safety shall satisfy the following requirements:
a)
The capacity for storing fuel oil and the supply system shall be sufficient to
meet demand for a specified period;
b)
Capability for start-up and operation under all circumstances and at all times;
c)
Auxiliary systems for the driving equipment, such as cooling systems, shall be
provided.
9.
The design shall also include features enabling the safe use of non-permanently
installed equipment to restore the required electrical power supply.
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Article 89. Performance of support systems and auxiliary systems
The
design of support systems and auxiliary systems shall ensure that the
performance of such systems is commensurate with the significance to safety of
the system or component that they support or serve.
Article 90. Auxiliary heat transfer systems
Systems
and components of the nuclear power plant that operate under operating states
and accident conditions shall be provided with auxiliary heat transfer systems,
the design of which shall ensure the isolation of non-essential portions.
Article 91. Process sampling system and post-accident sampling system
1.
Process sampling and post-accident sampling systems shall be provided to
promptly determine the concentration and activity of radionuclides present in
fluid treatment systems and in gaseous and liquid samples taken from systems or
the environment, under all operating states and accident conditions.
2.
Appropriate measures shall be provided to monitor activity in systems
containing liquids and gases that may be radioactively contaminated, and provisions
shall be made for sampling.
Article 92. Compressed air system
The
design basis of the compressed air system serving items important to safety
shall specify the quality, flow rate, and cleanliness of the air supplied.
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1.
Appropriate air conditioning, heating, cooling, and ventilation systems shall
be provided in areas of the nuclear power plant to maintain the environmental
conditions necessary for systems and components important to safety under all
plant states.
2.
Building ventilation systems shall be capable of purifying the air and shall
satisfy the following requirements:
a)
The dispersion of radioactive substances within the nuclear power plant shall
remain below the prescribed limits;
b)
The concentration and activity of radioactive substances in the air in areas to
which operating personnel are required to have access shall remain below the
prescribed limits;
c)
The concentration and activity of radioactive substances in the air within the
nuclear power plant shall remain below the prescribed limits and comply with
the ALARA principle;
d)
Ventilation of rooms containing inert or toxic gases shall not result in loss
of radiological control;
dd)
Gaseous radioactive releases into the environment shall be controlled below the
prescribed limits and in accordance with the ALARA principle.
3.
Lower pressure (a localized negative-pressure environment) shall be maintained
in areas with higher levels of radioactive contamination relative to areas with
lower levels of radioactive contamination where personnel have access.
Article 94. Fire prevention and firefighting systems
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2.
The installed fire prevention and firefighting systems shall be capable of
safely coping with fires postulated in the design.
3.
The fire suppression system shall be capable of automatic actuation at
appropriate locations. The design and arrangement of the fire suppression
system shall ensure that its failure, spurious actuation, or inadvertent
actuation does not significantly impair the capability of items important to
safety to perform their safety functions.
4.
The fire detection system shall be designed to provide operating personnel with
timely information on the location and spread of a fire.
5.
Fire detection and fire suppression systems necessary to protect the nuclear
power plant against fire hazards arising following a postulated initiating
event shall be capable of withstanding the effects of that event.
6.
Non-combustible or flame-retardant and heat-resistant materials shall be used
to the greatest extent practicable in the nuclear power plant, particularly in
areas such as the reactor bunker and the main control room.
Article 95. Lighting system
Adequate
lighting systems shall be provided in all operating areas of the nuclear power
plant under operating states and accident conditions.
Article 96. Lifting equipment
1.
Lifting equipment shall be provided for lifting and lowering items important to
safety or other items located near items important to safety.
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a)
Prevention of overloading during lifting and lowering operations;
b)
Prevention of load drops;
c)
Capability for the safe movement of the lifting equipment itself and the items
being lifted or lowered;
d)
Safety interlocks;
dd)
Seismic qualification where such equipment is used in areas containing items
important to safety.
Subsection 7. OTHER ENERGY CONVERSION SYSTEMS
Article 97. Steam supply system, feedwater system, and turbine generator
1.
The design of the steam supply system, feedwater system, and turbine generator
of the nuclear power plant shall ensure that the design limits of the reactor
coolant pressure boundary are not exceeded under operating states or accident
conditions.
2.
The steam supply system shall be equipped with appropriately rated and
qualified steam isolation valves capable of closing under all prescribed
conditions.
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4.
Turbine generators shall be provided with appropriate protective devices, such
as overspeed protection and vibration protection, and measures shall be
provided to minimize the potential impacts of turbine-generated missiles on
items important to safety.
Subsection 8. RADIOACTIVE WASTE TREATMENT SYSTEMS
Article 98. Radioactive waste treatment and control systems
1.
Systems for the treatment of solid and liquid radioactive waste shall be
provided at the nuclear power plant to maintain the quantity and concentration
of radioactive substances released below the permissible discharge limits and
in accordance with the ALARA principle.
2.
Systems and facilities shall be provided for the management and storage of
radioactive waste within the nuclear power plant area for a period appropriate
to the relevant treatment or disposal arrangements.
3.
The design shall incorporate features to facilitate the movement, transport,
handling, lifting, and packaging of radioactive waste. Consideration shall be
given to providing access to radioactive waste management and storage
facilities, as well as to the capability to perform lifting and packaging
operations for radioactive waste.
Article 99. Liquid and gaseous radioactive waste treatment and control
systems
1.
Systems for the treatment of liquid and gaseous radioactive waste shall be
provided at the nuclear power plant to maintain releases below the prescribed
discharge limits and as low as reasonably achievable.
2.
Liquid and gaseous radioactive waste shall be treated at the plant so that
doses received by the public as a result of releases into the environment
remain below the prescribed limits and as low as reasonably achievable.
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4.
Equipment for the retention of radionuclides contained in gaseous radioactive
waste shall have a sufficient retention efficiency to maintain radioactive
releases below the permissible discharge limits. The filtration system shall be
designed so that its efficiency can be tested, its performance and functions
can be regularly monitored throughout its service life, and filter elements can
be replaced while maintaining airflow through the system.
Subsection 9. FUEL HANDLING AND FUEL STORAGE SYSTEMS
Article 100. Fuel handling and fuel storage systems
1.
Fuel handling and storage systems shall be provided at the nuclear power plant
to ensure that the integrity and characteristics of the fuel are maintained at all
times during fuel handling and storage.
2.
The design shall incorporate features to facilitate the lifting, movement, and
handling of fresh fuel and spent nuclear fuel.
3.
The design shall ensure the prevention of significant damage to items important
to safety during the transport of fuel or in the event of a drop of fuel or a
fuel container.
4.
Fuel handling and storage systems, for both fresh fuel and irradiated fuel,
shall be designed to:
a)
Prevent criticality with an adequate safety margin, by physical means or
physical processes, with preference given to the use of geometrically safe
configurations, including under optimum neutron moderation conditions;
b)
Facilitate fuel inspection;
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d)
Prevent damage to fuel;
dd)
Prevent fuel from being dropped during transport;
e)
Provide identification markings for each fuel assembly;
g)
Provide radiation protection measures;
h)
Provide appropriate operating procedures and a system for nuclear material
accounting and control.
5.
In addition, the handling and storage system for irradiated fuel shall be
designed to:
a)
Ensure that the cooling capability of spent nuclear fuel is maintained under
all operating states and accident conditions;
b)
Prevent spent nuclear fuel from being dropped during transport;
c)
Avoid subjecting fuel rods or fuel assemblies to stresses exceeding permissible
limits;
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dd)
Safely store damaged or suspected damaged fuel rods or fuel assemblies;
e)
Control the concentration of dissolved neutron absorbers, where applicable, so
as to maintain a subcritical state;
g)
Facilitate maintenance and dismantling of the fuel handling and storage system;
h)
Facilitate decontamination where necessary;
i)
Provide adequate capacity for all fuel removed from the reactor core in
accordance with the anticipated core management plan;
k)
Facilitate removal of fuel from the storage area and preparation for transport
off-site.
6.
For a nuclear power plant using a water-filled storage pool for fuel storage,
the design shall prevent uncovering of fuel assemblies under all operating
states associated with the storage pool, so as to practically eliminate the
possibility of conditions that could lead to an early radioactive release or a
large radioactive release, while avoiding the creation of high-radiation areas
within the plant. The design shall ensure:
a)
Adequate fuel cooling capability is maintained;
b)
Features are provided to prevent uncovering of fuel assemblies in the event of
leakage or rupture of connected piping;
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d)
Features are provided to enable the safe use of non-permanently installed
equipment to ensure sufficient water inventory for long-term cooling of spent
nuclear fuel and adequate radiation shielding.
7.
The fuel storage system shall be equipped with monitoring and control means,
including:
a)
Monitoring and control of the storage pool temperature and water level under
relevant operating states and accident conditions;
b)
Monitoring of radioactivity in the water and in the air of the storage pool
area under relevant operating states and accident conditions;
c)
Monitoring and control of water chemistry parameters under operating states.
Subsection 10. RADIATION PROTECTION
Article 101. Design for radiation safety
1.
Design measures shall be provided to ensure that radiation doses to operating
personnel are maintained below the permissible dose limits and as low as
reasonably achievable, with full consideration given to the relevant dose
constraints.
2.
Radiation sources throughout the plant shall be fully and comprehensively
identified; associated radiation doses and radiological risks shall be
maintained in accordance with the ALARA principle; the integrity of the fuel
cladding shall be ensured; and the formation and transport of corrosion products
and activation products shall be controlled.
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4.
Measures shall be provided to prevent the release of radioactive substances,
radioactive waste, and radioactive contamination within the plant.
5.
The overall plant layout shall be arranged so as to ensure strict control of
personnel access to areas where there is a risk of radiation exposure or
radioactive contamination. Such layout, in combination with the ventilation
system, shall prevent or minimize radiation doses and the spread of
contamination.
6.
The plant shall be divided into controlled zones on the basis of anticipated
occupancy factors, radiation dose levels, and levels of radioactive
contamination under operating states and accident conditions. Appropriate
radiation shielding shall be provided to prevent or minimize radiation doses.
7.
The design and layout of the plant shall ensure that radiation doses to
personnel under normal operation, including fuel replacement and shuffling,
maintenance, and servicing, are maintained in accordance with the ALARA
principle, taking into account the use of necessary specialized equipment.
8.
Plant equipment requiring frequent maintenance or manual operation shall be
located in areas with low dose rates so as to reduce radiation exposure to
plant personnel.
9.
Facilities and equipment shall be provided for the decontamination of operating
personnel and plant equipment.
Article 102. Radiation monitoring means
1.
Equipment shall be provided to ensure monitoring of radiation levels under all
operating states and design basis accidents and, to the extent practicable,
under design extension conditions.
2.
Fixed dose-rate measuring equipment shall be provided to monitor radiation dose
rates at locations within the plant that are regularly accessed by personnel
and at locations where radiation levels vary under operating states and to
which personnel access is permitted only for specified periods.
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a)
Display radiation levels at necessary locations within the nuclear power plant
in the event of an accident;
b)
Provide sufficient information in the main control room and at control
locations to enable personnel to take intervention actions where necessary.
4.
Fixed monitoring equipment shall be provided to measure the activity of
radioactive substances in the air in areas:
a)
Where personnel regularly work;
b)
Where airborne radioactivity levels require the application of protective
measures;
c)
Where radioactive contamination may occur as a result of equipment failure or
other abnormal conditions.
5.
Means shall be provided to issue warning signals in the main control room and
at other appropriate locations when radionuclide concentrations exceeding
prescribed levels are detected. Monitoring equipment shall also be provided in
areas where radioactive contamination may occur as a result of equipment
failure or other anticipated operational occurrences.
6.
Fixed equipment and laboratory facilities shall be provided to promptly
determine selected radioactive concentrations in fluid treatment systems, as
well as in gaseous and liquid samples taken from plant systems or the
environment, under operating states and accident conditions.
7.
Fixed equipment shall be provided to monitor radioactive waste and potentially
contaminated waste before, during, and after discharge into the environment. Such
equipment shall be capable of automatically issuing an alarm when monitored
parameters exceed prescribed limits.
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9.
Facilities shall be provided for monitoring radiation doses and contamination
of operating personnel; procedures shall be established for assessing and
recording the accumulated radiation doses of workers over time.
10.
Measures shall be provided to assess radiation doses and other radiological
impacts, if any, in the area surrounding the plant through an environmental
monitoring program, including measurements of dose rates or activity
concentrations, with particular consideration given to:
a)
Exposure pathways to humans, including the food chain;
b)
Radiological impacts, if any, on the surrounding environment;
c)
The potential for accumulation and buildup of radioactive substances in the
environment;
d)
The potential existence of unauthorized radioactive discharge pathways.
Chapter IV
SAFETY
REQUIREMENTS FOR CONSTRUCTION OF NUCLEAR POWER PLANTS
Section 1. GENERAL REQUIREMENTS FOR CONSTRUCTION OF NUCLEAR POWER PLANTS
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1.
Where a nuclear power plant construction license has not yet been granted, and
the project owner wishes to commence advance fabrication of long-lead items,
the project owner shall:
a)
Notify the Vietnam Agency for Radiation and Nuclear Safety in writing;
b)
Ensure that the design of such items complies with the applicable requirements,
national standards, national technical regulations, international standards, and
foreign technical regulations, and has reached an appropriate level of
completion before construction commences.
2.
The establishment and assessment of construction methods and measures,
transportation methods, inspection, and testing relating to safety shall be
completed before such activities are carried out.
3.
The project owner, the Vietnam Agency for Radiation and Nuclear Safety,
contractors, and relevant parties shall maintain and promote a safety culture.
4.
The project owner shall establish environmental monitoring and environmental
protection measures and procedures to minimize environmental impacts and the
amount of waste generated from construction activities.
Article 104. Design management
1.
The design organization shall prepare a design schedule, including verification
of acceptance criteria and technical activities, in accordance with the
licensing process, and the project owner shall confirm such schedule before
construction.
2.
Before construction, the project owner and the construction organization shall
assess readiness and determine the level of design completion, the adequacy of
technical documentation, and any incomplete design-related matters.
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4.
During construction, the project owner shall manage and limit design changes
that may affect safety in accordance with clearly established procedures for
demonstrating the safety of the construction drawing design.
5.
The design and construction shall take into account the future decommissioning
of the nuclear power plant as prescribed in Article 38 of this Circular.
Article 105. Risk assessment
1.
Procedures, methods, and construction measures for the nuclear power plant
shall take into account internal and external hazards.
2.
The following activities shall be performed when conducting a risk assessment:
a)
Identify facilities at the site that contain hazardous materials, including
explosive, flammable, corrosive, toxic, or radioactive materials that are
handled, processed, or stored;
b)
Consider the quantities of materials presenting the greatest hazards, the
processes in which such materials are used, and their combined effects;
c)
Prepare a risk assessment inventory covering pipelines containing hazardous
materials and other hazard sources, such as temporary construction areas,
mines, and quarries where explosives are used and stored and which may
temporarily obstruct water flows, resulting in flooding or subsidence hazards
at the site.
Article 106. Assurance of radiation safety, nuclear safety, and
emergency response during construction of new nuclear power plant in area with
operating nuclear facility
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2.
The project owner and the operating organization shall implement preventive
measures to manage risks associated with construction activities.
3.
The project owner shall establish emergency response measures to ensure the
safety of personnel and the public in the event of an incident at the
construction site or an incident at the operating nuclear facility that affects
the construction site.
Section 2. RADIATION SAFETY AND NUCLEAR SAFETY SUPERVISION
Article 107. Contents of radiation safety and nuclear safety supervision
for construction activities
1.
The objectives and scope of supervision include:
a)
Radiation safety and nuclear safety supervision to verify that the project
owner fully complies with the conditions specified in the license and relevant
provisions of law;
b)
Radiation safety and nuclear safety supervision shall include, where necessary,
consideration of the activities of contractors providing services and products
to the project owner; supervision by the Vietnam Agency for Radiation and
Nuclear Safety shall not relieve the project owner of its primary
responsibility for ensuring radiation safety and nuclear safety.
2.
The contents of supervision include:
a)
Assessment of the quality management system of the project owner; the project
owner's control over principal contractors and subcontractors; and assessment
of the project owner's direct monitoring and supervision of construction work
and items relating to radiation safety and nuclear safety;
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c)
Examination of measurement and testing results and witnessing of activities
selected by the Vietnam Agency for Radiation and Nuclear Safety.
Article 108. Rights and responsibilities of Vietnam Agency for Radiation
and Nuclear Safety
1.
The Vietnam Agency for Radiation and Nuclear Safety shall have the following
rights:
a)
Establish hold points for the inspection and assessment of the results of
specific activities that are irreversible or would make independent inspection
or verification by the Vietnam Agency for Radiation and Nuclear Safety
difficult; progression to the subsequent stage shall only be permitted after
approval has been given by the Vietnam Agency for Radiation and Nuclear Safety;
b)
Require the project owner to prepare and implement remedial actions and
promptly provide information on non-conformities to the Vietnam Agency for
Radiation and Nuclear Safety where such non-conformities are identified.
2.
The Vietnam Agency for Radiation and Nuclear Safety shall have the following
responsibilities:
a)
Provide the project owner with early notification of the hold points
established under Point a Clause 1 of this Article so that the project owner
may consider and incorporate them into the construction plan and schedule;
b)
Allocate resources to ensure that radiation safety and nuclear safety
supervision is consistent, timely, and aligned with the progress of
construction activities;
c)
Develop a supervision program consistent with the project owner's construction
plan contained in the application dossier for the nuclear power plant
construction license;
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Section 3. MANAGEMENT OF CONSTRUCTION ACTIVITIES
Article 109. General requirements
Management
of construction activities shall include the following requirements:
1.
Planning, scheduling, and sequencing of work:
a)
Construction activities shall be planned, scheduled, and sequenced for
implementation. The plan shall include the following contents:
Construction
activities shall be divided into manageable work packages. Activities of a
complex nature, such as verification and validation of digital instrumentation
and control systems, qualification of items important to safety for
environmental conditions and seismic resistance, and other related activities,
shall be carefully planned.
The
anticipated sequence and duration of activities.
Resources
allocated to each activity.
b)
The construction organization shall continuously manage the construction
schedule throughout the construction process and communicate such schedule to
relevant parties. The construction project shall be fully integrated with the
procurement program, with particular attention paid to long-lead items;
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d)
Technical documents, drawings, plans, and schedules shall specify fabrication,
installation, inspection, and testing activities to be carried out at the
construction site and the conditions necessary for the proper performance of
such activities;
dd)
The construction organization shall ensure that on-site fabrication and
installation of items important to safety satisfy design and safety
requirements and comply with the applicable standards and technical regulations;
e)
On-site fabrication and installation activities shall be arranged at locations
where they do not adversely affect items important to safety in nearby areas or
other activities that may affect such items important to safety;
g)
The sequencing of construction activities shall ensure that previously
completed work is not adversely affected by subsequent construction activities.
2.
Technical requirements for procurement:
a)
Procurement technical requirements shall be developed in accordance with the
required lead time to ensure that items important to safety satisfy the
required design specifications and safety level;
b)
Procurement technical requirements for items important to safety shall fully
specify safety requirements relating to the technical characteristics of such
items, safety culture, and quality management;
c)
The safety classification of items important to safety shall be included in the
procurement technical requirements. Any change in safety classification shall
be notified to the supplier, and the impact of such change on equipment already
fabricated shall be assessed;
d)
Procurement documentation for items important to safety shall require a
manufacturing completion report. Such report shall include:
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Inspection
and testing results;
Non-conformity
reports;
Records
demonstrating that procurement has been carried out in accordance with
requirements;
Instructions
for preservation, installation, testing, and storage;
Operation
and maintenance manuals;
Operational
limits and conditions;
Personnel
training requirements;
As-built
drawings;
A
bill of materials containing a list of raw materials, parts, components, and
equipment, together with their corresponding quantities.
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Article 110. Fabrication and installation
1.
The quality management system of the project owner shall provide for review of
procurement documentation for items to be fabricated and installed to identify
the regulations, standards, technical regulations, and other requirements
applicable during fabrication and installation. Radiation safety, nuclear
safety, and other requirements specified in the procurement documentation shall
be appropriately reflected in fabrication drawings, technical specifications,
inspection and testing plans, procedures, and instructions.
2.
During the planning stage for fabrication and installation activities, the
project owner shall consider the following factors:
a)
The effects of the design on the fabrication process;
b)
Procurement of items affecting the construction schedule and long-lead items;
c)
Cleanliness control conditions and other environmental control measures
necessary to satisfy requirements and achieve the required quality;
d)
Equipment installation locations;
dd)
Requirements for transport, storage, packaging, and delivery;
e)
Application of new techniques in fabrication, installation, inspection, and
testing;
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h)
The need for inspection and testing specified by the design organization and
the Vietnam Agency for Radiation and Nuclear Safety, as well as inspections and
tests considered necessary by the manufacturer for quality control and to
ensure compliance with procedures;
i)
The need to establish, assess, and control new fabrication and installation
procedures;
k)
Complex or sensitive processes requiring extensive preparation, specialized
equipment, or special training;
l)
Compatibility of cleaning methods and cleaning materials with the items to be
cleaned.
3.
The project owner shall ensure that the quality management system of each
manufacturer includes identification and control of processes whose results
cannot be fully verified by subsequent inspection and testing and for which
process non-conformities may become apparent only after the item is placed into
service or operation.
4.
Special equipment, including tools, fixtures, molds, specialized inspection
equipment, and computational programs used during fabrication and installation,
shall be assessed and verified as suitable. Personnel using such equipment
shall be trained and shall clearly understand the limitations applicable to its
use.
5.
The project owner and the construction organization shall establish and
implement requirements and procedures for verifying the quality of fabrication
and installation of items important to safety.
6.
The technical capability of contractors shall be verified by the project owner
and the construction organization before procurement requirements are specified.
For new fabrication techniques and new equipment, the project owner and the
construction organization shall enhance supervision and inspection to verify
compliance with design requirements.
7.
All items important to safety shall be inspected and tested by the manufacturer
in accordance with the requirements specified in the procurement technical
documentation. Traceability of the history of each item and identification of
production batches shall be clearly addressed in the inspection and testing plan.
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9.
Prerequisites for construction work shall include:
a)
Where the construction drawing design is not yet available at the commencement
of construction, the design organization shall prepare and implement an action
plan to ensure that incomplete design documentation is completed in a timely
manner so as to ensure the quality of instructions, procedures, drawings, and
construction preparations;
b)
Contractors may perform work only after approval by the project owner and the
construction organization. Before carrying out each construction activity, the
contractor shall ensure that complete and up-to-date relevant information is
available, including the work schedule, instructions accompanied by drawings,
and appropriate consumable materials.
10.
Working and environmental conditions shall include:
a)
The construction organization shall monitor construction work and environmental
conditions to ensure that safety-related mechanical, electrical, and
instrumentation and control equipment and structures are protected against
damage caused by internal or external factors or contamination by dust, soil,
or foreign materials;
b)
The project owner shall specify permissible environmental conditions for
construction work, including temperature, pressure, humidity, precipitation,
dust, soil, airborne salt, wind, and electromagnetic conditions. The construction
organization shall periodically monitor environmental conditions to ensure that
they remain within permissible limits.
11.
Cleanliness and foreign material control shall include:
a)
When procuring items for installation, the project owner or the construction
organization shall ensure that cleanliness requirements are specified in the
procurement documentation so that items delivered to the construction site meet
the required level of cleanliness;
b)
The construction organization shall establish the necessary measures and
controls to protect items important to safety against internal and external
contamination by dust, soil, and foreign materials. Such measures shall include:
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Methods
for controlling environmental conditions;
Control
of personnel access. Where clean areas are used for such control, those clean
areas shall be clearly marked, and procedures and instructions shall be
established governing their use and maintenance;
Specification
and control of the types of chemicals and consumable materials permitted for
use;
Contingency
plans in the event that protective measures and controls are ineffective.
c)
Specialized cleaning procedures and methods shall be applied to systems
including hydraulic systems, instrumentation and control systems, and
lubricating oil lines, where internal surfaces generally cannot be inspected by
direct visual examination;
d)
Specific procedures shall be developed and implemented for cleaning by flushing
or rinsing. Such procedures shall include:
Inspection
of the actual flow path to ensure compliance with requirements concerning the
location, status, and condition of all components;
Tagging
and locking of important components to prevent inadvertent operation;
Inspection
of the internal cleanliness of all accessible components and piping;
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Sealing
of open ends of systems that have been cleaned;
Inspection
of the installation and removal of temporary equipment.
12.
Receipt, handling, transport, storage, preservation, and maintenance shall
include:
a)
Control of items and consumable materials
Items
and consumable materials shall be controlled through appropriate packaging,
transport, handling, receipt, and storage at all locations, including off-site
fabrication facilities, so as to prevent misuse, damage, deterioration, or loss
of identification.
b)
Handling
Where
handling activities may cause damage to items important to safety,
consideration shall be given to the use of equipment such as special-purpose
boxes, containers, protective devices, cranes, hoists, handling equipment, and
means of transport; operating personnel and personnel handling such equipment
shall be competent to perform their tasks; such equipment shall be used and
maintained in accordance with applicable regulations and national standards and
shall be subject to supervision by the construction organization to ensure the
safety of the items being handled.
c)
Transport
All
transport routes within and outside the construction site shall be planned with
appropriate protective measures for items important to safety. For the
transport of large or heavy components, all factors relating to the route shall
be adequately assessed to ensure that transport is carried out without causing
hazards, damage, or injury to persons, items, or property along the transport
route.
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Appropriate
storage shall be ensured in accordance with the requirements of the design
organization and the manufacturer to protect items important to safety before
installation and placement into service.
Storage
areas shall be established taking into account the following factors:
cleanliness and construction-site housekeeping practices; fire prevention and
firefighting requirements; requirements for protection of coatings,
preservation materials, covers, and protective tubing; prevention of mechanical
damage; control of environmental conditions; preventive maintenance; security;
physical and chemical characteristics of stored items; radiation protection;
and appropriate identification of radioactive sources.
The
construction organization shall conduct inspections where necessary to ensure
that the specified conditions are maintained and that any non-conformities are
appropriately addressed.
Items
important to safety and their components shall
be clearly identified by appropriate markings. Marking materials shall be
compatible with the items so as to ensure material preservation. The
construction organization shall implement physical protection at the
construction site to prevent unauthorized removal of items important to safety
that are in use or in storage.
Before
installing all items important to safety, the construction organization and
contractors shall inspect such items against the specified requirements and
implement remedial measures where necessary.
dd)
Preservation of installed items
Acceptable
limits for environmental and operating conditions that equipment may withstand
after installation shall be specified by the design organization and the
manufacturer. The construction organization shall monitor and control the
environmental conditions of installed items important to safety to protect them
against the effects of other work.
Contractors
and the construction organization shall ensure that the use of temporary
structures does not adversely affect items important to safety.
During
on-site or off-site testing, other items important to safety shall be isolated
or protected against unintended effects.
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Throughout
the construction stage, the project owner and the construction organization
shall ensure that items important to safety are appropriately maintained and
repaired so as to preserve the functions required by the design. The
maintenance and repair plan shall continue through the trial operation stage
until the operational maintenance program is established.
13.
Verification and testing of construction activities shall include:
a)
The project owner and the construction organization shall:
Develop
and agree upon procedures for verifying completion of construction activities
and handover of completed work.
Prepare
written test plans and acceptance criteria to enable independent assessment;
test results shall be compared against the acceptance criteria.
Ensure
that testing and verification of items important to safety are performed by a
competent independent organization.
Prepare
verification records demonstrating that the constructed items comply with the
requirements and satisfy the acceptance criteria, including those specified in
the application dossier for the license.
b)
The verification records prescribed in Point a of this Clause shall include the
following contents:
Identification
of structures, systems, and components;
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Date
and time of verification;
Name
and organization of the verification personnel;
Calibration
instruments or equipment used;
Test
results and comparison with the acceptance criteria;
List
of outstanding deficiencies and the remedial action plan;
List
of outstanding work and the implementation plan;
Confirmation
that the required records and documents are complete.
c)
The use of sealed radioactive sources and radiation equipment in activities
such as radiography, use of radiation measuring equipment, or material analysis
shall comply with radiation safety requirements.
Article 111. Construction processes at construction site
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a)
When items are received at the construction site, an initial inspection shall
be conducted to ensure that the items conform to the purchase order and have
not sustained external damage during transport;
b)
After receipt, the construction organization shall inspect the items to ensure
that they comply with the technical specifications before acceptance and use.
Such inspection shall include confirmation of the following:
The
item has the correct configuration;
Identification
and markings are complete;
Fabrication
and installation documentation is available as required;
Inspection
records and certificates are traceable to the inspected item and demonstrate
compliance with requirements;
Protective
covers and seals are intact;
Coatings
and preservation materials are undamaged;
There
is no mechanical damage;
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Inert
gas or desiccant, where applicable, has not been adversely affected;
Non-conformities
identified or arising that require repair at the construction site have been
recorded;
Necessary
tests of hardware characteristics have been performed;
Storage
is controlled to prevent incorrect installation or use.
2.
Management of interactions with an operating nuclear facility shall include:
a)
The project owner shall conduct a safety assessment during construction, taking
into account all hazards posed by adjacent facilities to the construction site
and by the construction site to adjacent facilities, including interdependencies
among safety systems; the potential radioactive contamination impacts from
operating nuclear facilities on the construction site and other risks shall
also be assessed, including the impacts of the total environmental discharges
from all facilities within the same site area;
b)
The responsibilities of the project owner and the construction organization
shall be agreed upon before construction activities commence at the site. The
parties shall cooperate closely and exchange information and shall implement
all necessary measures to ensure that operating nuclear facilities are capable
of safe operation throughout the construction process;
c)
For construction of a nuclear power plant adjacent to an operating nuclear
facility, or where buildings or support services are intended to be shared, the
following boundaries shall be clearly identified: radiation protection
boundaries, physical boundaries, system boundaries, access boundaries, and
cleanliness control area boundaries.
Where
services from the operating nuclear facility, such as water, electricity,
compressed air, fire prevention and firefighting services, and emergency
medical services, are used, cooperation mechanisms shall be clearly defined and
understood by the construction organization so as to avoid adversely affecting
the safety of the operating nuclear facilities.
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Procedures
shall be established to ensure that the organization licensed to operate the
nuclear facility approves any proposed change in the status of shared buildings
or services before such change is implemented by the construction organization.
3.
On-site fabrication and installation shall include the following:
a)
Temporary equipment and tools used in fabrication, installation, inspection,
and testing shall be controlled and documented;
b)
Waste materials and remaining consumable materials used or generated at the
construction site during construction shall be appropriately collected and
treated by the contractor upon completion of the work.
Chapter V
REQUIREMENTS FOR
NUCLEAR POWER PLANT SAFETY ANALYSIS
Section 1. GENERAL REQUIREMENTS FOR NUCLEAR POWER PLANT SAFETY ANALYSIS
METHODS
Article 112. Scope of safety analysis
1.
Nuclear power plant safety analysis shall be conducted using both deterministic
safety analysis and probabilistic safety analysis methods to assess the safety
level of the plant under different states and operating modes. In conducting
deterministic safety analysis and probabilistic safety analysis, applicable
national standards, national technical regulations, international standards,
and foreign standards or technical regulations shall be applied.
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3.
Nuclear power plant safety analysis shall determine the frequencies and
consequences associated with anticipated operational occurrences and accident
conditions.
4.
Safety analysis shall include events arising from internal hazards, external
hazards, and processes that may damage barriers for the confinement of
radioactive substances or increase the risk of radioactive leakage into the
environment. External hazards having a low frequency of occurrence but
potentially leading to reactor core melting shall be taken into account in
severe accident analysis.
5.
The selection of events for analysis and their subsequent progression shall be
based on a systematic and logical approach. Adequate justification shall be
provided for the identification of all accident scenarios relevant to safety.
6.
Specific requirements for the scope of probabilistic safety analysis shall
include:
a)
Level 1 probabilistic safety analysis shall be performed to identify the
frequency of events that may lead to reactor core melting; estimate the core
melt frequency; and assess the strengths and weaknesses of safety systems and
operating procedures for preventing reactor core melting;
b)
Level 2 probabilistic safety analysis shall be performed to identify potential
radioactive release pathways in the event of a severe accident, estimate the
magnitude and frequency of such releases, and assess the adequacy of measures
for preventing and mitigating radioactive releases into the environment;
c)
Analyses shall be performed for the reactor and the spent nuclear fuel storage
pool under all operating modes and plant states of the nuclear power plant;
d)
Analysis of initiating events shall include events internal to the plant, human
errors, and internal and external hazards.
Article 113. Responsibility for safety analysis
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2.
Safety analysis shall be performed by persons having appropriate qualifications
and experience and a thorough understanding of all aspects of nuclear power
plant safety analysis.
Article 114. Results of deterministic safety analysis and probabilistic
safety analysis
1.
The results of deterministic safety analysis shall include a comparison of the
analysis results with the acceptance criteria prescribed in Section 2 of
Chapter V of this Circular and the following:
a)
Confirmation of the adequacy of the design basis for all items important to
safety, as well as the adequacy of the operational limits and conditions and
the required actions of operating personnel;
b)
Confirmation that the postulated initiating events are appropriate to the
characteristics of the site and the design of the nuclear power plant;
c)
Justification that management of anticipated operational occurrences and design
basis accidents through actuation of safety systems complies with the
acceptance criteria;
d)
Justification that management of design extension conditions through the use of
safety features unaffected by the accident sequence complies with the
acceptance criteria.
2.
The results of probabilistic safety analysis shall include comparison of the
analysis results with the acceptance criteria prescribed in Section 2 of
Chapter V of this Circular and the following:
a)
Justification that no postulated initiating event makes an excessively large
contribution to the overall risk or a significant contribution to the
uncertainty of the analysis results;
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c)
Justification that cliff-edge effects affecting safety as a result of
variations in input parameters are prevented.
Article 115. Specific requirements for deterministic safety analysis
1.
In conducting deterministic safety analysis, an adequate safety margin shall be
maintained between the calculated values of important parameters and threshold
values that lead to radioactive releases, including where the best-estimate
approach is used.
2.
Deterministic safety analysis for design purposes shall ensure conservatism,
with reasonable consideration given to model uncertainties, except for the
analysis of design extension conditions.
3.
The selection of calculation data and assumptions shall take into account uncertainties
relating to the following factors:
a)
Initial plant operating conditions;
b)
Operability of safety systems;
c)
Actions of operating personnel;
d)
Availability of grid power for the actuation of safety systems.
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a)
Relevant acceptance criteria and limiting physical parameters shall be
identified;
b)
Initial conditions and boundary conditions shall be selected conservatively as
appropriate for each acceptance criterion. Sensitivity analysis shall be
performed, where necessary, to justify such selection;
c)
Plant systems and equipment not designed to operate under particular states
shall be assumed to have failed or to operate in a manner that aggravates the
initiating event, unless their operability can be demonstrated with high
reliability;
d)
In the analysis of a design basis accident, the most severe single failure in
the operation of a safety system shall be assumed.
Article 116. Uncertainty analysis and sensitivity analysis
1.
The results of safety analysis shall include the results of uncertainty
analysis and sensitivity analysis.
2.
Uncertainties shall be quantified, and the method used to address them shall be
clearly justified, taking into account the origin and nature of the uncertain
factors.
Article 117. Computational programs used in safety analysis
1.
Computational methods, computational programs, and computational models used in
safety analysis shall be verified and validated.
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3.
During model validation, uncertainties and errors in the model, as well as
limitations of the model and database, shall be identified, together with the
manner in which such factors are taken into account in the safety analysis.
4.
It shall be ensured that persons using a computational program have sufficient
competence and experience in applying that program to the safety analysis of a
nuclear power plant.
Article 118. Use of operating experience in safety analysis
1.
Data on operating experience that may be used for safety analysis shall be
collected and assessed, including consideration of anticipated operational
occurrences and accidents that have occurred during operation at similar
nuclear power plants. The scope of data collected shall be consistent with the
graded approach.
2.
Operating experience data shall include:
a)
Records of anticipated operational occurrences and accidents occurring during
operation;
b)
Errors by operating personnel;
c)
Capability of safety systems to perform their functions;
d)
Reliability of items important to safety;
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e)
Generation of radioactive waste.
3.
Throughout the operating lifetime of the nuclear power plant, data shall be
collected on the basis of plant safety performance indicators. Operating
experience data shall be appropriately used, together with advances in science
and technology relating to safety analysis methods and relevant research
results, to update the safety analysis results and assess the management
system.
Section 2. ACCEPTANCE CRITERIA FOR SAFETY ANALYSIS
Article 119. Requirements for establishing acceptance criteria
1.
Acceptance criteria shall be established for all operating states and accident
conditions. Such criteria shall ensure that an adequate level of defense in
depth is maintained and that no unacceptable level of hazard is posed to humans
or the environment.
2.
Acceptance criteria shall be established on the basis of the following factors:
a)
Characteristics of the postulated initiating event, particularly its frequency
of occurrence;
b)
Reactor technology type;
c)
Actual plant conditions, particularly the capability for self-supplied
electrical power;
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3.
Acceptance criteria for each operating state or accident condition shall
include:
a)
General criteria relating to the consequences of radioactive releases;
b)
Specific criteria relating to the integrity of barriers against radioactive
releases, including fuel pellets, fuel rod cladding, the reactor coolant
pressure boundary, and the reactor bunker.
Article 120. Acceptance criteria for deterministic safety analysis of
consequences of radioactive releases
1.
The effective dose to plant personnel and the public under operating states
throughout the plant lifecycle shall comply with the ALARA principle. The dose
limit for a member of the public shall be less than 1 mSv/year.
2.
All design basis accidents shall not result in significant radiological impacts
at or beyond the site and shall not require off-site protective actions. The
effective dose to a member of the public resulting from a design basis
accident, excluding intake of radioactive substances through ingestion, shall
not exceed 5 mSv.
3.
Design extension conditions that may lead to a large radioactive release into
the environment shall be practically eliminated. For other design extension
conditions, measures shall be provided to limit releases after a specified
period and within a specified area so as to allow sufficient time for implementation
of protective measures for the public.
4.
Releases of radioactive substances from a severe accident shall not be
permitted to result in the following consequences:
a)
Acute radiation injury to the health of the public in the area surrounding the nuclear
power plant;
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c)
Release of Cs-137 into the environment with radioactivity exceeding 100 TBq;
d)
Three months after the occurrence of the accident, deposition of a combination of
radionuclides other than Cs-137 causing a hazard greater than that resulting
from a Cs-137 release at the limit specified in Point c of this Clause.
Article 121. General acceptance criteria for deterministic safety
analysis
1.
Design features, automatic actuation capability of safety systems, together
with the required actions of operating personnel, shall be sufficiently
effective to ensure that:
a)
The plant state is not aggravated;
b)
No additional single failure occurs;
c)
There is no loss of operability of a safety system or safety feature necessary
to mitigate accident consequences.
2.
A system used to mitigate the consequences of an event shall be capable of
withstanding the maximum loads, stresses, and environmental conditions
corresponding to the event being analyzed.
3.
Pressure in the primary and secondary systems shall not exceed the design
limits applicable to each specified plant state.
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5.
For all design basis accidents, the geometry of the reactor core shall be
maintained so as to preserve core coolability.
6.
A design basis accident shall not result in temperature, pressure, pressure
differential, or other loads acting on the reactor bunker exceeding the values
used as the design basis.
Article 122. Specific acceptance criteria for deterministic safety
analysis
1.
Specific acceptance criteria shall be established to ensure adequate safety
margins and the integrity of barriers against radioactive releases.
Consideration shall be given to applying the following specific acceptance
criteria:
a)
Criteria relating to the integrity of the fuel pellet structure, including:
maximum temperature within the fuel pellet; maximum radially averaged fuel
enthalpy, together with their dependence on burnup and fuel composition;
b)
Criteria relating to the integrity of the fuel rod cladding, including: minimum
Departure from Nucleate Boiling Ratio (DNBR) for a pressurized water reactor,
or minimum Critical Power Ratio (CPR) for a boiling water reactor; maximum fuel
rod cladding temperature; and maximum local oxidation of the fuel rod cladding;
c)
Criteria relating to the integrity of the reactor core, including: degree of
subcriticality; maximum amount of hydrogen generated by oxidation of the fuel
rod cladding; and maximum number of damaged fuel rods and fuel assemblies;
d)
Acceptance criteria relating to the integrity of the reactor cooling system
(primary circuit), including: maximum coolant temperature and pressure; changes
in temperature, pressure, and stress imposed on the reactor coolant pressure
boundary; and absence of brittle or ductile cracking or fracture resulting from
postulated defects in the reactor vessel;
dd)
Acceptance criteria relating to the integrity of the secondary circuit,
including: maximum temperature and pressure and changes in coolant pressure and
temperature in secondary-circuit equipment;
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2.
For a postulated initiating event occurring during reactor shutdown or an event
resulting in degradation of the integrity of protective barriers, more
stringent acceptance criteria shall be applied to prevent the uncovering of
fuel assemblies.
Article 123. Acceptance criteria for probabilistic safety analysis
1.
For Level 1 probabilistic safety analysis, the total core melt frequency shall
be less than 10-5 per reactor-year.
2.
For Level 2 probabilistic safety analysis, the total frequency of a release of
Cs-137 exceeding 100 TBq into the environment shall be less than 10-6
per reactor-year.
Chapter VI
REQUIREMENTS FOR
QUALITY MANAGEMENT SYSTEM FOR NUCLEAR POWER PLANT SITE SURVEY, SITE ASSESSMENT,
DESIGN, CONSTRUCTION, TRIAL OPERATION, OPERATION, AND OPERATIONAL TERMINATION
Section 1. GENERAL REQUIREMENTS FOR QUALITY MANAGEMENT SYSTEM APPLICABLE
TO ORGANIZATIONS PARTICIPATING IN NUCLEAR POWER PLANT SITE SURVEY, SITE
ASSESSMENT, DESIGN, CONSTRUCTION, TRIAL OPERATION, OPERATION, AND OPERATIONAL
TERMINATION
Article 124. Responsibilities of head of organization for quality
management system
1.
Bear primary responsibility for the quality management system, even where
individuals or organizational units have been assigned to perform related tasks.
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3.
Issue and disseminate safety policies, organize their implementation, maintain
them, and periodically review them.
4.
Establish the organizational structure, functions, and tasks of the
organization; clearly define responsibilities and assign tasks to each
individual or organizational unit when carrying out site survey, site
assessment, design, construction, trial operation, operation, and operational
termination of the nuclear power plant; and ensure clarity as to when, how, and
by whom decisions are made within the quality management system.
5.
Establish reporting mechanisms for implementation of the quality management
system, including the contents, methods, timing, and authority for reporting.
6.
Demonstrate commitment to maintaining implementation of the quality management
system, including organizing training and supervision of activities important
to safety; developing and disseminating values and standards of behavior
relating to safety; and ensuring that all individuals understand, accept, and
implement such safety values and standards.
7.
Ensure that all individuals within the organization participate in the
implementation and continuous improvement of the quality management system.
Article 125. Responsibilities of head of organization for organization’s
objectives, strategies, plans, and targets
1.
Establish and maintain objectives, strategies, plans, and targets consistent
with the organization’s safety policy.
2.
Ensure that objectives are measurable at all levels of the organization.
3.
Periodically review the implementation of objectives and plans, assess the
extent to which targets have been achieved, and implement remedial measures
where deviations occur.
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1.
Identify relevant parties and establish appropriate cooperation mechanisms,
ensuring that safety is prioritized in all cooperation activities.
2.
Establish procedures and cooperation plans with relevant parties on the basis
of the identified cooperation mechanisms, including:
a)
Methods for periodic cooperation and provision of information to relevant
parties concerning radiation risks;
b)
Mechanisms for timely and effective communication with relevant parties in
unforeseen situations or where changes occur during activities;
c)
Methods for disseminating necessary safety-related information to relevant
parties;
d)
Methods for considering safety-related assessments and requirements of relevant
parties in the decision-making process.
3.
Establish specific requirements on the basis of identified and agreed
expectations of relevant parties that are to be met, and disseminate such
requirements throughout the management system.
4.
Ensure that safety is always prioritized where the expectations of relevant
parties conflict or may affect the level of safety.
5.
Promote transparency, compliance, and ethics in cooperation and interaction
with relevant parties.
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7.
Use the assessment results as input for improvement of the quality management
system.
Article 127. Quality management system and integration of quality
management system
1.
The quality management system shall be established, implemented, assessed, and
continuously improved, and shall be appropriate for and support the achievement
of the organization’s safety and quality objectives; it shall fully integrate
quality management requirements and describe planned and systematic actions to
ensure that such requirements are met.
2.
The quality management system shall be applied to enhance safety and promote
safety culture by:
a)
Integrating all elements necessary for the management of safety within the
organization and its activities;
b)
Describing requirements and activities relating to quality management and
safety assurance;
c)
Describing planned and systematic actions necessary to ensure that all requirements
are met;
d)
Ensuring that safety is always taken into account in decision-making and is not
adversely affected by decisions.
3.
The quality management system shall include arrangements for resolving
conflicts that may arise during decision-making or work performance. Potential
impacts of security measures on safety, as well as impacts of safety measures
on security, shall be identified and resolved without compromising safety or
security.
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5.
The quality management system shall establish mechanisms for conducting an
independent assessment before important safety-related decisions are made and
shall specify requirements for the independence of the assessment and the
competence of the assessors.
6.
The quality management system shall include the following contents:
a)
Applicable provisions of law; applicable national standards, national technical
regulations, international standards, and foreign standards and technical
regulations; relevant safety guides; formal agreements with relevant parties;
and international practices;
b)
Organizational structure; objectives, strategies, plans, and targets of the
organization;
c)
Responsibilities and authorities of the head of the organization, management,
functional units, and individuals;
d)
Processes necessary for the performance of work, including mechanisms for
internal interaction and cooperation and interaction and cooperation with
relevant parties so as to ensure quality throughout the process;
dd)
Integration of the quality management system into the organization’s overall
management system;
e)
Application of the graded approach, commensurate with the importance of the
process and factors affecting safety;
g)
Documentation of the entire quality management system, control of documents and
records, and assurance of traceability, retention, and confidentiality of
information as prescribed;
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i)
Management of processes, activities, and changes; establishment of reporting
mechanisms, including contents, timing, methods, and persons responsible for
reporting;
k)
Management of contractors, suppliers, and the supply chain relating to items,
products, and services affecting quality and safety;
l)
Safety policy and the promotion and maintenance of safety culture;
m)
Measurement, periodic assessment, and continuous improvement of the quality
management system;
n)
Internal control, external independent assessment, and management of
non-conformities.
Article 128. Application of graded approach
1.
The quality management system shall be developed and applied in accordance with
the graded approach, commensurate with management requirements and the characteristics
of the activities.
2.
The criteria used to classify the extent to which the quality management system
is developed and applied shall be specified in the organization’s quality
management system.
3.
In determining the criteria prescribed in Clause 2 of this Article, full
consideration shall be given to the following factors:
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b)
Potential hazards and risks relating to safety and quality;
c)
Potential safety consequences in the event of errors, unexpected events, or
incomplete implementation of planned activities.
Article 129. Documentation of quality management system
1.
The quality management system shall be issued in documented form. Documentation
of the quality management system shall be controlled and shall be usable,
legible, clearly identifiable, and readily accessible at the point of use.
2.
Documentation of the quality management system shall include the contents
prescribed in Clause 6 Article 127 of this Circular.
3.
All individuals responsible for preparing, reviewing, revising, and approving
documents shall be competent to perform their assigned tasks and shall be
provided with appropriate information as the basis for their opinions or
decisions.
4.
Document revisions shall be controlled, reviewed, and tracked. Revised
documents shall be re-approved.
5.
Requirements for record retention shall be specified in the quality management
system and shall be controlled. All records shall be legible, complete,
identifiable, and readily retrievable.
6.
Retention periods for records and related test materials and specimens shall be
determined in accordance with provisions of law and the organization’s
knowledge management obligations. Storage media shall ensure that records
remain legible throughout the specified retention period.
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1.
The head of an organization participating in site survey, site assessment,
design, construction, trial operation, operation, and operational termination
of a nuclear power plant shall determine the resources necessary for the
organization to carry out its activities in compliance with safety requirements
and shall ensure the provision of such resources. Resources shall include
personnel (in terms of number and competence), infrastructure, working
environment, knowledge and information, suppliers, materials, and financial
resources.
2.
The head of the organization shall have the following responsibilities:
a)
Establish the quality management system and operating processes to ensure that
the organization possesses, or maintains access to, all capabilities and
resources necessary to conduct its activities and discharge its
responsibilities for ensuring quality at each stage of the lifecycle of the
nuclear power plant, as well as in the event of an emergency response;
b)
Determine the capabilities and resources that the organization must retain or
develop internally and those that may be provided externally to ensure quality;
c)
Establish specific competence requirements, provide appropriate training and
advanced training, and assess the effectiveness of such training and advanced
training to ensure that all personnel at all levels remain capable of performing
their assigned tasks safely and effectively.
3.
Capabilities to be maintained internally within the organization shall include:
a)
Leadership capability at all management levels;
b)
Capability to promote and maintain safety culture;
c)
Expertise necessary to understand the technical, human, and organizational
aspects relating to the nuclear power plant so as to ensure safety and quality.
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5.
The organization’s knowledge and information shall be managed as a resource.
Article 131. Management of processes, activities, and changes
1.
Processes and activities shall be developed and effectively managed to achieve
the organization’s objectives without adversely affecting safety.
2.
Processes shall be issued in documented form and shall be consistent with the
organization’s existing documentation, accompanied by records demonstrating the
results achieved by such processes.
3.
The sequence of processes and their interactions shall be specified to ensure
that safety is not adversely affected. Interrelated processes shall ensure
effective interaction, with particular consideration given to interactions
among processes within the organization and interactions between the
organization’s processes and those of external service providers.
4.
New processes or amendments to existing processes shall be reviewed, approved,
and implemented in a manner that does not adversely affect safety. Amended
processes shall be consistent with the organization’s objectives, plans, and
targets.
5.
Processes shall include all required inspection, testing, verification, where
applicable, and validation activities, together with the associated acceptance
criteria and responsibilities for implementation, and shall identify the time
and stage at which such activities are to be performed.
6.
Each process or activity that may affect safety or quality shall be performed
under controlled conditions through compliance with approved procedures,
instructions, and drawings. Such procedures, instructions, and drawings shall
be approved before first use and periodically reviewed to ensure their adequacy
and effectiveness. Individuals performing such activities shall participate in
the periodic review process.
Article 132. Management of contractors and supply chain
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2.
The organization shall establish cooperation mechanisms with suppliers,
contractors, and other supply parties to specify, monitor, and manage the
provision to the organization of items, products, and services that may affect
safety.
3.
The organization shall have a clear and comprehensive understanding of the
products or services supplied. The organization shall maintain internal
capability to specify the scope and standards of the products or services
required and subsequently assess whether the supplied products or services
satisfy safety and quality requirements.
4.
The organization shall retain responsibility for safety when outsourcing
processes and when receiving items, products, or services from the supply chain.
5.
The organization shall establish mechanisms to ensure that suppliers of items,
products, and services important to safety comply with safety requirements and
meet the organization’s expectations regarding safety performance.
6.
Competence assessment, selection, qualification, procurement, and supervision
of the supply chain shall be incorporated into the quality management system.
Article 133. Promotion and maintenance of safety culture
1.
All individuals within the organization, from the head of the organization
downward, shall be responsible for implementing and promoting safety culture.
The management and leadership system for safety shall be established with the
objective of promoting and maintaining safety culture.
2.
The head of the organization and all levels of management shall support and
facilitate the promotion of the following:
a)
A common understanding of safety and safety culture, including awareness of
radiation risks and hazards associated with the work and working environment;
understanding of the impacts of such radiation risks and hazards on safety; and
collective commitment to safety by groups and individuals;
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c)
Promotion of a culture of cooperation, trust, and open communication;
d)
Timely reporting of issues relating to technical, human, and organizational
factors and reporting of deficiencies in structures, systems, and components to
avoid degradation of safety, including timely acknowledgement and feedback
regarding actions taken;
dd)
Measures encouraging a questioning attitude and a learning attitude at all
levels of the organization, and preventing complacency with respect to safety;
e)
Application of methods and measures to enhance safety and maintain safety
culture through a systemic approach that fully considers interactions among
technical, human, and organizational factors;
g)
Safety-oriented decision-making for all activities;
h)
Enhancement of communication, sharing, and integration between safety culture
and security culture.
Article 134. Measurement, periodic assessment, and improvement of
quality management system
1.
The effectiveness of the quality management system shall be monitored and
measured to confirm the organization’s capability to achieve the intended
results and to identify areas for improvement.
2.
All processes shall be periodically assessed for their effectiveness and
capability to ensure safety and quality.
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4.
Self-assessment and independent assessment of the quality management system,
management responsibilities for safety, and safety culture shall be conducted
regularly to assess the effectiveness of the management system and identify areas
requiring improvement. The impacts on safety of lessons learned and any
significant changes that arise shall be analyzed. The results of
self-assessment and independent assessment shall be communicated to all levels
of the organization and used to enhance the effectiveness of the quality
management system, improve leadership capability for safety, and promote a
learning attitude within the organization.
5.
The organization shall assign responsibility for conducting independent
assessments of the quality management system, management responsibilities for
safety, and safety culture. Organizations, units, whether internal or external,
and individuals assigned such responsibilities shall be granted sufficient
authority to perform their tasks and shall have direct access to the head of
the organization. Individuals conducting independent assessments shall not
assess activities under their direct managerial responsibility.
6.
The head of the organization shall periodically review the quality management
system to confirm that it remains appropriate, effective, and capable of
supporting the organization in achieving its established objectives. Such
review shall take into account changes in legal and technical requirements and
changes within the organization.
7.
The quality management system shall include timely assessment and application
of the following:
a)
Lessons learned from activities and events occurring within and outside the
organization;
b)
Technical advances and research and development results;
c)
Lessons learned from the identification of good practices.
8.
The organization shall establish mechanisms and arrangements for learning from
achieved successes and strengths to support organizational development and
continuous improvement.
Article 135. Internal control, external independent assessment, and
management of non-conformities
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a)
Internal control means assessment activities carried out by senior management,
management at all levels, and individuals within the organization themselves to
assess work performance and the improvement of the organization’s safety
culture.
b)
Internal control activities shall be carried out in accordance with the
principles prescribed in Clause 4 Article 134 of this Circular.
2.
Independent assessment:
a)
Independent assessment means assessment activities regularly conducted by a
unit or individual having no direct responsibility for the work being assessed,
to evaluate the effectiveness of processes in meeting and achieving the
organization’s objectives, strategies, and plans; determine the adequacy of
work performance and leadership capability; assess the organization’s safety
culture and leadership for safety; monitor product quality; and identify
opportunities for improvement. The unit or individual conducting the
independent assessment shall have qualifications and professional competence
appropriate to the field of work being assessed.
b)
Independent assessment activities shall be conducted in accordance with the
principles prescribed in Clauses 4 and 5 Article 134 of this Circular.
3.
Management of non-conformities:
a)
Non-conformities include abnormal conditions and events or failures to meet
specific requirements, including: deviations from approved parameters or
procedures; delivery or procurement of items or services that fail to satisfy
requirements; individuals failing to follow work instructions; incomplete
documentation or documentation containing inaccurate information; and
insufficient training of individuals to perform their assigned tasks.
b)
Management of non-conformities shall be carried out in accordance with the
principles prescribed in Clause 3 Article 134 of this Circular.
Section 2. REQUIREMENTS FOR QUALITY MANAGEMENT SYSTEM FOR NUCLEAR POWER
PLANT SITE SURVEY AND SITE ASSESSMENT
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1.
An organization conducting site survey and site assessment for a nuclear power
plant shall develop and implement a quality management system in compliance
with the requirements prescribed in Section 1 of Chapter VI of this Circular,
the safety requirements for nuclear power plant site survey and site assessment
prescribed by the law on atomic energy, and other relevant laws, and shall
appropriately apply relevant international guidance on quality management
systems or the requirements and guidance on quality management systems of the
technology-supplying country when specifying requirements for the quality
management system.
2.
The quality management system shall ensure that the collection, analysis,
verification, and management of site data are carried out objectively,
consistently, and in a verifiable manner to ensure the completeness and
reliability of the site assessment results.
3.
The quality management system shall fully take into account the requirements
and safety guidance relevant to nuclear power plant site survey and site
assessment.
Article 137. Application of graded approach
When
applying the graded approach in the quality management system for site survey
and site assessment activities, the organization shall consider:
1.
The ultimate intended use of the knowledge and data obtained from the site
assessment activities, particularly the extent to which such knowledge and data
may affect safety.
2.
The capability to demonstrate, verify, or reproduce the results.
3.
The scale and technical complexity of site survey and site assessment
activities, including the use of new models or the extension of existing models.
4.
The managerial complexity of the activities and the degree of participation and
cooperation among different disciplines, units, and organizations, as well as
the degree of interdependence among tasks.
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6.
Requirements for and intended use of the results.
Section 3. REQUIREMENTS FOR QUALITY MANAGEMENT SYSTEM FOR NUCLEAR POWER
PLANT DESIGN
Article 138. Requirements for quality management system for nuclear
power plant design
1.
The nuclear power plant design organization shall develop and implement a
quality management system in compliance with the requirements prescribed in
Section 1 of Chapter VI of this Circular, the safety requirements for nuclear
power plant design prescribed by the law on atomic energy, and other relevant
laws, and shall appropriately apply relevant international guidance on quality
management systems or the requirements and guidance on quality management
systems of the technology-supplying country when specifying requirements for
the quality management system.
2.
The quality management system shall ensure that:
a)
It includes measures to ensure the design quality of each structure, system,
and component and of the overall nuclear power plant design, ensuring that the
design complies with safety requirements and the applicable national standards,
national technical regulations, international standards, and foreign standards
and technical regulations;
b)
It includes the processes necessary for preparation, appraisal, and approval of
the design; identification and remediation of design errors or deficiencies;
verification of design completeness; control of design changes; and assurance
of the design configuration;
c)
All safety requirements established for the plant design are considered and
implemented at every stage of the design process and are satisfied in the final
design;
d)
The completeness, consistency, verifiability, and traceability of all design
documentation, data, and records are ensured.
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When
applying the graded approach in the quality management system, the design
organization shall consider:
1.
The scope and level of detail of the design analysis.
2.
The extent of design review and approval.
3.
The extent of design verification.
4.
Requirements for control of design changes.
5.
The level of detail of design records and the record retention period.
6.
The need to perform alternative calculations.
7.
The need for design output testing.
8.
The need for design qualification testing.
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Article 140. Requirements for quality management system for nuclear
power plant construction activities
1.
The nuclear power plant construction organization shall develop and implement a
quality management system in compliance with the requirements prescribed in
Section 1 of Chapter VI of this Circular, the safety requirements for nuclear
power plant construction prescribed by the law on atomic energy, and other
relevant laws, and shall appropriately apply relevant international guidance on
quality management systems or the requirements and guidance on quality
management systems of the technology-supplying country when specifying
requirements for the quality management system.
2.
The quality management system for nuclear power plant construction activities
shall ensure that:
a)
It fully describes the mechanisms and organizational measures for managing,
implementing, and assessing construction activities; and ensures that
construction complies with the design, provisions of law, and applicable
national standards, national technical regulations, international standards,
and foreign standards and technical regulations;
b)
It identifies the organizational structure, responsibilities, and cooperation
mechanisms of organizations participating in construction, including suppliers,
the construction organization, contractors, the design organization, the site
assessment organization, and other relevant organizations;
c)
Any changes arising during construction are controlled, their impacts are
assessed, and they are incorporated into the design and construction records.
Article 141. Responsibilities of project owner
1.
Ensure that the nuclear power plant is constructed in compliance with all
requirements for radiation safety, nuclear safety, and nuclear security.
2.
Establish and maintain the capabilities necessary to control all activities
within the scope of the granted license.
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a)
Developing and implementing a management system covering construction
activities to ensure that the required level of quality for safety is achieved.
The project owner shall establish a plan for supervising construction
activities relating to items important to safety, including independent
inspection, product quality monitoring, hold points, and field inspections;
b)
Identifying and maintaining core capabilities in supervision activities and
contracting processes;
c)
Receiving and retaining ownership of safety records, particularly information
provided by contractors, the design organization, or the construction
organization;
d)
Conducting internal and external assessments of contractors’ quality management
systems based on the graded approach. The project owner shall conduct
supervision to verify that contractor activities comply with all relevant
safety requirements, both technically and with respect to the quality
management system;
dd)
Inspecting, testing, and verifying items important to safety;
e)
Establishing and implementing procedures for handling design changes,
non-conformities, and initiating events;
g)
Reporting design changes to items important to safety, non-conformities, and
initiating events as required by the Vietnam Agency for Radiation and Nuclear
Safety;
h)
Ensuring that safety-related records are established and retained, particularly
records relating to each stage of the nuclear power plant lifecycle and ageing
management;
i)
Preparing for the trial operation program, taking into account tests performed
during construction;
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Article 142. Activities of construction organization
The
construction organization shall carry out the following activities:
1.
Develop cooperation arrangements between the construction organization and
relevant organizations and incorporate them into procurement documentation or
contracts.
2.
Control and supervise contractors participating in activities of safety
significance, both on-site and off-site, including fabrication and installation
activities.
3.
Ensure that the construction organization and on-site contractors are arranged
and controlled within designated areas and are provided with necessary site
services, information, and instructions relating to radiation safety, nuclear
safety, and nuclear security requirements.
4.
Identify common construction activities and develop and maintain guidance on the
application of standardized procedures, instructions, and good practices.
5.
Develop safety-related work procedures for personnel of the construction
organization and contractors.
6.
Monitor the implementation of safety policies and activities of all personnel
to ensure compliance with requirements relating to quality and safety.
7.
Plan and monitor the progress of construction project work and ensure
compliance with the established schedule.
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9.
Ensure that appropriate preservation and maintenance measures are implemented
for installed equipment that may be damaged during construction.
10.
Conduct inspections of structures, systems, and components to collect baseline
data for comparison with inspections conducted during subsequent operation.
11.
Organize and control the handover of work and records between work groups.
12.
Procure, install, and maintain material specimens for long-term monitoring and
analysis programs relating to material ageing.
13.
Ensure that complete documentation is prepared to demonstrate careful
assessment, compliance with requirements, and the recording and reporting of
non-conformities and remedial measures through process inspection and
supervision of construction activities.
Article 143. Application of graded approach
1.
The construction organization shall apply the graded approach to the
implementation of safety requirements on the basis of the significance to
safety of each item, service, or process.
2.
When applying the graded approach, the construction organization shall consider:
a)
Whether special construction processes, such as non-destructive examination,
personnel, and equipment, meet the applicable requirements;
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c)
The significance to safety of equipment, materials, processes, records, and
other documentation;
d)
The complexity of the activity;
dd)
The level of traceability;
e)
The level of process control and the need for hold points or witness points;
g)
Procurement activities;
h)
Other relevant risks.
Article 144. Requirements for implementation of construction processes
1.
Construction processes shall be identified through review of the scope of
construction activities derived from design requirements for structures,
systems and components, procurement documents, drawings, construction work
plans, and schedules.
2.
Review of the requirements, documents, drawings, plans, and schedules shall
identify the specified on-site fabrication, installation, inspection, and
testing activities.
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4.
The principal activities of personnel of the construction organization shall
include, at a minimum:
a)
Controlling and supervising suppliers on-site and off-site;
b)
Ensuring control of on-site suppliers within designated areas and the
appropriate provision of necessary services, information, and instructions
relating to industrial safety requirements;
c)
Developing safety-related work procedures, including industrial safety
procedures, for issuance to personnel of both the construction organization and
contractors, and ensuring that the construction organization and contractors at
the construction site comply with applicable requirements;
d)
Monitoring industrial safety policies and activities of all personnel at the
construction site to ensure compliance with provisions of law;
dd)
Planning and monitoring work progress to comply with the construction program,
including cooperation in activities of suppliers responsible for the
construction of structures, systems, and components;
e)
Ensuring that work performed by suppliers complies with procedures,
requirements, and drawings and satisfies specified quality requirements; and
ensuring that inspections and tests at supplier facilities conform to the
relevant inspection and testing plans and supervision schedules;
g)
Performing maintenance on equipment that may deteriorate during construction,
such as drying electrical equipment and preserving critical surfaces that may
be subject to corrosion;
h)
Conducting initial inspections of systems or components that will require
inspection during operation;
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k)
Arranging controlled handover of completed work from the current supplier to
another supplier or to the construction organization;
l)
Collecting baseline data for comparison during operational inspections;
m)
Ensuring compliance with the law in documentation relating to the work.
Article 145. Resources for construction activities
1.
The project owner and the construction organization shall ensure that a
sufficient number of suitable personnel with the qualifications and experience
required for the construction project are available, and shall establish
procedures to ensure initial competence assessment and maintenance of personnel
competence.
2.
Resources for construction activities shall be estimated, planned, and ensured
for the construction of items important to safety, in accordance with the
project schedule, particularly for long-lead items.
3.
Personnel of the project owner and the operating organization who will
participate in the trial operation, operation, and maintenance of the nuclear
power plant should, where practicable, participate in construction activities
to receive practical training and gain additional experience in operation,
maintenance, and technical support.
Article 146. Control and supervision of contractors
1.
Assessment and selection of contractors:
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b)
The graded approach shall be applied in establishing criteria for the
assessment and selection of contractors and suppliers of materials, products,
and services.
2.
Supervision of contractors:
a)
The extent of supervision by the project owner and the construction
organization over contractor activities shall be determined on the basis of the
graded approach. The extent of supervision shall depend on factors including:
The
significance to safety of the item or service;
The
experience of the project owner and the contractor with the product or service;
The
presence of any first-of-a-kind features;
The
complexity of the work or service;
The
expertise required to perform the work or service;
Available
evidence demonstrating that the required quality has been achieved;
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Requirements
of the Vietnam Agency for Radiation and Nuclear Safety.
b)
Before commencing activities following the conclusion of a subcontract, the
contractor shall demonstrate to the construction organization and the project
owner that it fully understands all requirements relevant to such activities.
Specific safety requirements applicable to the subcontract shall include:
Safety
requirements applicable to the item being fabricated or constructed;
Cooperation
arrangements;
Information
exchange arrangements;
Records
and information to be submitted, including non-conformity reports and evidence
demonstrating that the completed item complies with safety and quality
requirements;
The
quality management system.
c)
The construction organization shall be informed of all subcontractors selected
by the contractor. Any new subcontractor selected after the initial agreement
shall be notified to the construction organization and shall be approved before
commencement of the relevant work;
d)
The construction organization shall organize periodic meetings with contractors
and subcontractors to review and ensure implementation of the requirements of
the construction organization;
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Article 147. Requirements for handover and transfer of responsibility
1.
The construction organization shall establish provisions for controlling and
cooperating in the handover of completed works from one supplier to another and
from the construction organization to the organization responsible for trial
operation of the nuclear facility to maintain the integrity of the completed
works.
2.
Transfer of responsibility during construction:
a)
The construction organization shall establish and issue appropriate provisions
and procedures for controlling and cooperating in the handover of completed
items from one contractor to another to maintain the integrity of the completed
items. Provisions and procedures for access control to items important to
safety and work areas shall also be documented and applied during the transfer
process. Such provisions and procedures shall be approved by the project owner
before being used in the construction project.
b)
Where items important to safety and work areas are transferred between groups
within the construction organization or between contractors, the two parties
concerned shall jointly inspect the transferred items and related documentation
at the handover location. The parties shall agree on the configuration status
of the item, the work area, and identified non-conformities.
c)
After the transfer, any remaining work or remedial measures that remain the
responsibility of the transferring party may only be performed with the
approval of the receiving party.
3.
Transfer to the trial operation stage:
a)
The project owner shall ensure the establishment and implementation of
provisions for controlling and cooperating in the handover from the
construction stage to the trial operation stage. Such provisions shall include:
Documentation
relating to the items being transferred shall be reviewed by the construction
organization for completeness and accuracy. Any incomplete items or
non-conformities shall be identified, and the status of such items shall be
clearly established to ensure that they cannot adversely affect safety during
trial operation activities. Termination points defining the boundaries of the
systems and equipment being transferred, or the transferred portions thereof,
shall be clearly identified in the handover documentation;
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Incomplete
or non-conforming items shall be identified and assessed to ensure that they do
not adversely affect safety during trial operation;
Outstanding
work shall be agreed upon, planned, and scheduled;
The
termination points defining the boundaries of systems and equipment being
handed over, or the portions thereof being handed over, together with the
corresponding required configuration, shall be clearly identified in the
handover documentation;
The
items being handed over and the related records and documents shall be
inspected;
Approved
as-built drawings shall be handed over together with complete and accurate
configuration information for the nuclear power plant;
All
transferred structures, systems, and components shall be marked or tagged as
prescribed;
All
temporary equipment shall be clearly identified;
The
transfer of responsibility shall be fully documented.
b)
To ensure maintainability and ageing management when the nuclear power plant
enters operation, the handover documentation shall contain sufficient technical
detail to enable the operating organization to identify components and prepare
replacement components for maintenance activities. The project owner shall be
provided with copies of all other relevant information in addition to the
handover documentation.
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Article 148. Planning of construction activities
1.
The construction organization shall prepare a plan for construction activities,
including the following:
a)
Activities to be performed;
b)
Planned sequence and duration of implementation;
c)
Resources allocated to each activity.
2.
The construction organization shall retain responsibility for cooperation and
overall planning for construction of the nuclear facility.
3.
Suppliers shall be responsible for detailed planning of the work to be
performed and, where necessary, shall obtain approval of such plans from the
construction organization.
4.
When planning on-site fabrication, installation, inspection, and testing of
structures, systems, and components important to safety, the following
requirements shall be taken into account:
a)
Identification, preparation, and control of procedures and work instructions;
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c)
Competent personnel;
d)
Hold points for inspection and supervision by the Vietnam Agency for Radiation
and Nuclear Safety
dd)
Environmental considerations;
e)
Confirmation, at the end of construction, of the records to be transferred to
the organization responsible for trial operation or the operating organization
for retention throughout the operating lifetime of the facility.
Article 149. Kick-off meeting
1.
Following the conclusion of subcontracts during the construction stage, the
supplier and the construction organization shall hold a kick-off meeting to
confirm that the supplier is fully aware of the requirements of the
construction organization, including:
a)
Interaction mechanisms;
b)
Communication methods;
c)
Documents and information to be submitted;
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dd)
Site security;
e)
Site training;
g)
Industrial safety, particularly in the use of non-destructive examination and
during construction activities;
h)
The quality management system;
i)
Supervision and inspection of sub-suppliers.
The
meeting shall also finalize the measures to be implemented by the supplier to
satisfy these requirements.
Article 150. Control of design information
1.
The construction organization shall establish information channels and
communication mechanisms to ensure the issuance of design information among the
organizations participating in the design. Before issuance, the construction
organization shall ensure that the information to be issued reflects actual
conditions at the site. Particular attention shall be paid to design
information required at any off-site fabrication facility.
2.
The construction organization shall establish a procedure for handling queries
from suppliers concerning issued design information. If a query relates to
operational safety, it shall be referred to the principal design organization
for response.
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4.
The project owner shall ensure traceability of items important to safety from
the initial design stage through construction, trial operation, and subsequent
stages of the nuclear power plant life cycle, whereby:
a)
Appropriate procedures shall be established to ensure the traceability of
records for items important to safety from the design, construction, trial
operation, operation, and decommissioning stages of the nuclear power plant,
and such records shall be retained as prescribed;
b)
Records to be retained to ensure traceability shall include: as-built drawings;
details of fabrication and installation processes; inspection reports; site
traceability records, including identification markings and tags; construction
and test records; design calculations; documentation of design changes and
non-conformities; and details of equipment qualification;
c)
Traceability requirements for items important to safety shall apply to all
equipment, materials, processes, records, and other documents to ensure that
such important items are constructed to the appropriate quality level
corresponding to their safety classification.
Article 151. Housekeeping during construction
1.
Construction processes shall ensure that structures, systems, and components
are constructed in accordance with the specified cleanliness requirements,
including the establishment of necessary controls to protect mechanical,
electrical, and control equipment from internal and external contamination by
dust, dirt, or other foreign material. Particular attention shall be paid to
preventing foreign material from entering piping systems by controlling
openings in such systems during installation.
2.
Piping systems shall be flushed to meet cleanliness requirements before being
placed into operation.
3.
When procuring items, cleanliness requirements shall be included in procurement
documentation to ensure that items delivered to the site have an acceptable
level of cleanliness.
4.
For the preservation of items under construction or installation, measures for
site housekeeping, cleaning, and preservation shall be implemented, including:
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b)
Measures for controlling environmental conditions and personnel access. Where
clean areas are used for such control, they shall be clearly identified, and
procedures or instructions governing their use and maintenance shall be
established.
Article 152. Control of construction items
1.
Construction items shall be controlled from receipt through storage, handling,
and use to prevent misuse, mishandling, damage, deterioration, or loss of
identification. Items delivered to the construction site shall, where
appropriate, be visually inspected before unloading to confirm that they have
not been damaged.
2.
After items have been received, inspections shall be conducted to ensure that
the relevant requirements are satisfied, including:
a)
The item has the correct configuration;
b)
Identification and markings are complete;
c)
Required manufacturing documentation is available;
d)
Protective covers and seals are intact;
dd)
Coatings and preservation materials are undamaged;
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g)
The applicable cleanliness standards are met;
h)
The inert gas blanket and the condition of desiccants, if any, have not been
adversely affected;
i)
Required tests of hardware characteristics have been performed.
Article 153. Storage
1.
Storage areas shall be arranged as prescribed to segregate and protect items
before installation and use, together with provisions specifying storage
methods and conditions to prevent corrosion, contamination, deterioration, and
damage.
2.
Storage areas shall be established and controlled, taking into account the
following aspects:
a)
Access control;
b)
Housekeeping and site management;
c)
Fire prevention and firefighting requirements;
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dd)
Protection requirements relating to coatings, preservation materials, covers,
and protective sleeves;
e)
Prevention of physical damage;
g)
Removal from and return to storage areas;
h)
Environmental control, such as temperature and humidity;
i)
Preventive maintenance;
k)
Security;
l)
Items having a shelf life or service-life limitation;
m)
Physical and chemical characteristics of items;
n)
Safety classifications.
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Article 154. Handling of construction items
b)
Dimensions;
c)
Susceptibility to impact damage;
d)
Surface condition;
dd)
Specified handling points;
e)
Orientation;
g)
Lifting equipment and any tests required for such equipment;
h)
Susceptibility to deterioration due to electrostatic discharge;
i)
Preservation of coatings;
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2.
The use of items such as special crates, containers, protective equipment,
lifting equipment, handling devices, and means of transport shall be considered
where handling activities could cause damage. Operators and persons handling
all such equipment shall be competent. Equipment used for handling items shall
be operated and maintained in accordance with applicable regulations and
national standards and technical regulations.
3.
Items procured by the construction organization for provision to contractors
shall be stored and preserved to ensure that such items do not deteriorate and
remain capable of performing their intended design functions. Contractors shall
transfer the records for such items as part of the handover documentation,
where appropriate.
Article 155. Verification of construction work
1.
The construction organization shall establish verification methods and
schedules specifying the extent of inspection and verification required.
2.
Before requesting acceptance of an item or service, the supplier shall verify
that all procurement requirements have been satisfied. Acceptance by the
purchaser shall neither relieve the supplier of the responsibility to provide
products that are fit for their intended purpose nor preclude subsequent
rejection of such products.
3.
Construction activities performed by contractors and suppliers shall be carried
out on the basis of inspection and test plans. Depending on the complexity of
the work and the significance to safety of the system, hold points and witness
points shall be established in the inspection and test plans for assessing the
relevant activities and verifying acceptability.
4.
The process for verification of construction work shall include mechanisms for
verifying completion of construction and installation activities. Verification
shall be formally documented by means of check sheets or equivalent means to
confirm that structures, systems, and components have been constructed and
installed as specified.
5.
A check sheet shall include the following:
a)
Identification of the structures, systems, or components;
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c)
Time of inspection;
d)
Special tools or calibrated equipment used;
dd)
A list of missing items and outstanding items or work;
e)
Confirmation of the availability and completeness of documents and records;
g)
Confirmation by all parties that the inspection has been performed.
Article 156. Measurement, assessment, and improvement
1.
The project owner and the construction organization shall assess the quality
management system as follows:
a)
Periodically assess the effectiveness of implementation of the quality
management system, including assessment of progress to identify at an early
stage problems relating to performance, planning, and resource provision, and
consider measures to avoid adverse effects on the quality of work, including:
early implementation of solutions, adjustment of human resources, revision of
schedules, and renegotiation of contracts;
b)
Establish and implement an independent assessment program to consider issues
relating to project management capability.
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a)
Establish a system for collecting information on all identified non-conformities,
recording and addressing such non-conformities as prescribed, and ensuring that
all individuals participating in construction activities are responsible for
identifying and reporting non-conformities when they occur;
b)
Non-conformities having significant implications for safety shall be considered
as initiating events and addressed through a remedial and preventive action
program in accordance with the graded approach. It shall also be ensured that
the accumulation of multiple minor non-conformities does not result in changes
having significant implications for safety without being identified and fully
assessed;
c)
Retain records of remedial and preventive measures implemented to address
non-conformities and monitor the effectiveness of the process for addressing
such non-conformities;
d)
Refrain from delaying preventive measures solely because remedial measures
require a long period to complete. All outstanding issues relating to
non-conformities and remedial measures shall be tracked until completion, and
records shall be retained and relevant parties informed.
3.
The project owner and the construction organization shall implement
construction experience feedback as follows:
a)
Collect construction experience and good practices from nuclear power plants
under construction and from the construction of nuclear power plants and other
power plants; disseminate lessons learned to improve quality and safety; ensure
that such information is disseminated to relevant parties in the nuclear
industry; and establish mechanisms for the systematic and timely sharing of
construction experience;
b)
Proactively share safety-related experience with relevant domestic and
international parties.
Section 5. REQUIREMENTS FOR QUALITY MANAGEMENT SYSTEM FOR TRIAL
OPERATION OF NUCLEAR POWER UNITS
Article 157. Requirements for quality management system for trial
operation of nuclear power units
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2.
The quality management system shall be established on the basis of
consideration of general management system requirements and technical
requirements relating to trial operation of nuclear power units, and shall
fully describe the processes, organizational structure, and arrangements
necessary for organizing, performing, and controlling trial operation
activities.
3.
Trial operation stages, tests, and technical milestones shall be fully
described and implemented in a consistent manner. Trial operation procedures,
assessment processes, individual tests, and integrated tests shall ensure
consistency and conformity with the established technical requirements.
4.
Test procedures and related documentation shall be fully compiled and made
ready for transfer to the operating organization of the nuclear power plant.
System handover shall ensure completeness of records, traceability, and
participation of representatives of the operating organization.
Article 158. Application of graded approach during trial operation stage
The
organization conducting trial operation shall apply the graded approach, taking
into consideration:
1.
The level of detail of and need for inspection plans and test plans during
trial operation.
2.
The level of traceability required for data, records, and test results during
trial operation.
3.
The level of process control and the need to establish hold points and witness
points during trial operation.
4.
Qualification requirements for trial operation processes and procedures and for
personnel performing such processes.
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6.
Compilation of current test procedures and all related documentation for the
preparation of the handover records for the operating organization.
7.
Requirements for the handover of systems that have undergone trial operation to
representatives of the operating organization.
8.
Participation of operating personnel in trial operation stages.
9.
Requirements relating to final acceptance testing before completion of the
trial operation stage.
Section 6. REQUIREMENTS FOR QUALITY MANAGEMENT SYSTEM FOR OPERATION OF
NUCLEAR POWER PLANTS
Article 159. Requirements for quality management system for operation of
nuclear power plants
1.
The operating organization of a nuclear power plant shall establish and
implement a quality management system in compliance with the requirements
specified in Section 1 Chapter VI of this Circular, safety requirements for
operation of nuclear power plants prescribed by the law on atomic energy, and
other relevant laws, and shall appropriately apply relevant international
guidance on quality management systems or requirements and guidance on quality
management systems of the technology-supplying country when specifying
requirements for the quality management system.
2.
The quality management system for the operation of a nuclear power plant shall
be established on the basis of consideration of general management system
requirements and technical requirements relating to nuclear power plant
operation; shall cover the processes necessary to ensure safe and stable
operation in accordance with the design; and shall fully take into account
relevant safety guidance on nuclear power plant operation.
Article 160. Application of graded approach during operation stage
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a)
The need for and level of detail of operating instructions;
b)
The types of installed equipment requiring calibration;
c)
The level of reporting and authority for addressing non-conformities and
remedial actions;
d)
The need to maintain formal written shift operating logs;
e)
Testing, monitoring, and inspection activities during operation;
g)
Equipment required for monitoring and controlling plant conditions;
h)
Control requirements for the storage and records of relevant spare parts;
i)
The need to analyze the history of items within the facility;
k)
The need to perform condition monitoring;
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m)
The need to collect, assess, and use feedback from internal and external
operating experience.
Section 7. REQUIREMENTS FOR QUALITY MANAGEMENT SYSTEM FOR
DECOMMISSIONING OF NUCLEAR POWER PLANTS
Article 161. Requirements for quality management system for
decommissioning of nuclear power plants
1.
The organization carrying out the decommissioning of a nuclear power plant
shall establish and implement a quality management system in compliance with
the requirements specified in Section 1 Chapter VI of this Circular, safety
requirements for decommissioning of nuclear power plants prescribed by the law
on atomic energy, and other relevant laws, and shall appropriately apply
relevant international guidance on quality management systems or requirements
and guidance on quality management systems of the technology-supplying country
when specifying requirements for the quality management system.
2.
The quality management system shall be established on the basis of
consideration of general management system requirements, requirements relating
to radiation safety, nuclear safety, and management of waste generated from the
decommissioning process, and shall fully describe the processes, organizational
structure, and arrangements necessary for managing, performing, and controlling
decommissioning activities.
Article 162. Application of graded approach during decommissioning stage
When
applying the graded approach during the decommissioning stage, the organization
shall consider:
1.
The need for and level of detail of decommissioning documents, including the
decommissioning plan and related documents.
2.
Requirements for management of waste generated during decommissioning,
including collection, classification, treatment, and safe storage.
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4.
The type and level of detail of training programs for personnel participating
in nuclear power plant decommissioning activities.
5.
The necessary controls for dismantling the plant, removing equipment, and demolishing
structures. Safety shall be ensured throughout the implementation process.
Chapter VII
IMPLEMENTATION
PROVISIONS
Article 163. Implementation provisions
1.
This Circular shall enter into force from the date of its signing and abrogate
the following Circulars:
a)
Circular No. 13/2009/TT-BKHCN dated May 20, 2009 of the Minister of Science and
Technology of Vietnam;
b)
Circular No. 28/2011/TT-BKHCN dated October 28, 2011 of the Minister of Science
and Technology of Vietnam
c)
Circular No. 30/2012/TT-BKHCN dated December 28, 2012 of the Minister of
Science and Technology of Vietnam;
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dd)
Circular No. 20/2025/TT-BKHCN dated October 6, 2025 of the Minister of Science
and Technology of Vietnam.
2.
The Director of the Vietnam Agency for Radiation and Nuclear Safety, directors
of agencies and units under the Ministry of Science and Technology of Vietnam,
project owners, nuclear power plant operating organizations, and relevant
agencies, organizations, and individuals shall implement this Circular.
3.
Difficulties arising during the implementation of this Circular shall be
promptly reported to the Ministry of Science and Technology of Vietnam for guidance,
consideration, or resolution./.
MINISTER
Nguyen Manh Hung