ABSTRACT: includes station blackout (SBO), loss of all


Electricity is
considered to be one of the key factors for development of a country. To meet
up the electricity demand, the first unit of Rooppur nuclear power plant (RNPP)
is being constructed in the western region of the country. The NPP unit is of
VVER-1200 model, which is a generation three plus pressurized water reactor
(PWR) with increased service life. The design and construction of the NPP meet
all the regulatory requirements of International Atomic Energy Agency (IAEA)
and Bangladesh Atomic Energy Regulatory Authority (BAERA). This model
incorporates defense in depth concept. This paper reports on the safety
measures of RNPP during regular operation as well as on emergency and severe
operating conditions. During regular operation of the reactor, radiation safety
has been ensured by several safety barriers. Additionally, emergency operating
procedures (EOP) are introduced in case of occurrence of any event outside
design consideration, which includes station blackout (SBO), loss of all
secondary feed water (LSFW) etc. On the other hand, severe accident management
guidelines (SAMG) are introduced in case of failure of EOPs. These strategies
have been studied in this paper.  

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The reactor used in Rooppur nuclear power plant is of
model VVER-1200/523 of AES-2006 type. This is a pressurized water reactor,
where light water is used both as coolant and moderator. Each unit of VVER-1200
has 3200 MW of thermal capacity resulting to total 1200 MW of gross electrical
capacity. The primary circuit pressure at the core outlet and temperature at
the outlet are 16.2 Mpa and 328.9ºc respectively.  The first unit of the reactor is planned to
start electricity generation by 2023. The first
two units of VVER-1200 are being built at Leningrad Nuclear
Power Plant II and Novovoronezh Nuclear
Power Plant II.  Novovoronezh II-1 was connected to the grid and
started commercial operation on 27 February 2017. The VVER-1200 (or NPP-2006 or
AES-2006) is an evolution of the VVER-1000 being offered for domestic and
export use. VVER-1200 is the latest version of VVER reactors. VVER stands for Vodo-Vodyanoi
Energetichesky Reaktor, which means water water energetic reactor in
English. Power output ranges from 70 to 1200 MWe, with designs of up to 1700 MWe in
development. VVER power stations are in operation, under construction or
planned in the following countries: Armenia, Belarus, Bulgaria, China, Czech
Republic, Egypt, Finland, Germany, Jordan, Hungary, India, Iran, Slovakia,
Ukraine, Russia and Vietnam.






The reactor core
consists of 163 fuel assemblies (FAs). Each FA contains 312 fuel rods. The fuel
rod is cladded with zirconium alloy tube. There is up to 121 rod cluster
control assemblies (RCCAs) to suppress quick chain reaction. The RCCAs are also
used to maintain power to an assigned level and to its level-to-level transition.
RCCAs are controlled by drive mechanism. This drive mechanism is placed at the
top head of the reactor. The average burnup of unloaded fuel is up to 60 MWD/kg
U. Annually 42 fresh FAs are loaded into the core for the basic fuel cycle.


Primary-side Components:

§  Reactor Pressure

vertical reactor pressure vessel houses the core, control rods and in-core
instrumentation sensors. The RCCA nozzles are installed at the top head
nozzles. A concrete cavity is built around the reactor for biological &
thermal shielding. A cooling system is also installed at the concrete cavity.
The reactor design is depicted in figure 1.

§  Steam Generator:

steam generator with supports comprises of the following components: horizontal
steam generator, steam header, support, shock absorbers, one- and two chamber
surge tank, embedded components for supports and shock. The horizontal steam
generator is a single0vessel heat exchange apparatus of horizontal type with
heat transfer surface and the following components:

1.       a vessel with different-purpose nozzles;

2.       a heat-exchange bundle with fastener and
spacer components;

3.       primary coolant collectors;

4.       feedwater supply and distribution systems;

5.       emergency feedwater supply and distribution

6.       distribution perforated plate;

7.       submerged perforated plate;

8.       chemicals feeder;

The primary coolant collector has a variable diameter,
where the maximum diameter is 1176 mm at wall thickness 171 mm. Its inner
surface has protective corrosion-resistant cladding.

1: Reactor of VVER 1200 (AES-2006) Model




§  Reactor Coolant Pump (RCP):

RCP is designed for
primary coolant circulation. It has an additional function of providing coolant
circulation under any loss-of power accident.


§  Main Coolant
Pipeline (MCP):

The main coolant pipeline connects the reactor, steam
generators and reactor coolant pump. The main coolant pipeline comprises four
circulation loops. Each loop is divided into three tube sections. The reactor
outlet nozzle-to-SG inlet collector pipeline section is the hot leg. The cold
leg comprises two sections: the SG outlet collector-to-RCP set suction nozzles
pipeline and RCP set discharge nozzle-to-reactor inlet nozzle pipeline section.


§  Pressurizer:

The pressurizer is
a vertically positioned pressurized cylindrical vessel with elliptic bottoms
installed on a cylindrical support.



The time between
refueling is 12 months. The standard fuel cycle length is 4 years. MOX fuel can
be used as alternative fuel in the reactor.



In the containment
close to the reactor cavity, there is a spent fuel pond. The spent fuel is
placed in the pond for a time, then these are transferred to the spent fuel
tank. The spent fuel can be stored there for 10 years. 



K-1200-6.8/50 turbine will be installed in
the unit. This is a single-shaft five-cylinder set and comprises a double-flow
high pressure cylinder (HPC) and four double-flow low pressure cylinders (LPC).
Without the generator the turbine is ~ 52.3 m long, and with the generator it
is ~74.5 m long. The schematic thermal diagram of the turbine plant comprises
four stages of low pressure heaters, a deaerator, two stages of high pressure
heaters. The accepted thermal diagram of the turbine plant is designed for a
regenerative heatup of the main condensate and feedwater under normal operating
conditions. The schematic diagram of the turbine is 2LPC+HPC+2LPC.




Main Feedwater

Under normal operating conditions the
system is designed to perform the functions of maintaining the level in steam
generators and providing them with feedwater. T he system provides the control
of water supply to the steam generator under all the design-basis conditions
(including start-up and shutdown modes). The system of the main feedwater
pipelines comprises four main feedwater pumps (4×25%), one backup feedwater
pump (1×25%), valves and pipelines. The quality analysis of the system shows
that it meets the imposed safety requirements of the regulatory documents and
performs its functions under all the conditions that require its operation.
There are no deviations from the technical documentation and standards.


Auxiliary Feedwater System:

The auxiliary feedwater system is designed:

§  to provide steam generators with feedwater
under the conditions of normal operation (commissioning hydraulic tests, main
steam line heat up, shutdown to the state of hot standby, cooldown);

§  to provide steam generators with feedwater
under anticipated operational occurrences in case of a shutdown and subsequent
cooldown, under the conditions of loss of normal heat removal through the
secondary side (through the turbine condenser) as well as in case of loss power
to the station auxiliaries;

§  to provide steam generators with feedwater
under failures that are accompanied with limitations in water supply from the
system of main feedwater at the Unit power operation;

§  to provide the pre-startup deaeration of
feedwater (with the secondary-side deaerator);

§  to provide the post-cooling of steam
generators in water-to-water conditions at the Unit repair shutdown.

The auxiliary feedwater system comprises
two pumps, valves and pipelines.




The safety system configuration follows the
following principles:

§  single failure principle;

§  redundancy principle;

§  diversity principle;

§  principle of physical separation;

§  protection against the operator’s errors;

§  RP inherent safety principle.


Active and Passive Safety System:

active and passive safety system varies from different models of VVER-1200.
There the safety feature of V-1200 (V-392M) is presented. VVER-1200/ 592 will
incorporate developed safety features than V-1200 (V-392M).

active safety systems are:

§  Emergency steam generator cooldown systems;

§  Emergency gas removal system;

§  Emergency boron injection system;

§  System of the primary circuit emergency and planned cooldown and
spent fuel pond cooling;

§  Main steamline isolation system.


The passive safety systems are as

§  Emergency core cooling system, passive part;

§  Emergency core passive cooling systems;

§  Systems of passive heat removal


Defense in Depth:

The defense-in-depth concept based on
the application of a system of physical barriers to the release of the ionizing
radiation and radioactive products into the environment, and the systems of
engineering and organizational measures to protect the barriers and maintain
their efficiency, as well as to protect the personnel, the population and

The systems of
physical barriers are as follow:

§  A fuel matrix

§  A fuel rod cladding

§  Reactor coolant circuit boundary

§  Reactor plant enclosure

§  Biological shielding

The defense in depth system incorporates 5 levels of

The activities in this level includes NPP siting, establishment of control area
and off-site surveillance area. A design elaboration is performed to ensure
well-developed inherent safety of RP. The required quality of systems (components)
of NPP is ensured. The operation of the NPP must be monitored by timely
detecting the defects, taking preventive measures, replacement of equipment
with expired service life and arrangement of efficient of recording for work
and monitoring results. Another important activity is NPP personnel recruitment
according to their level of qualification for activities under normal operating
conditions, anticipated operational occurrences including the pre-accident
situations and accidents.

These activities are focus on the prevention of design basis accidents (DBAs)
by normal operating systems. This includes detection of deviations from normal
operating conditions and management of the deviated situation.

These activities aim at the prevention of beyond design basis accidents by
safety systems. The activities include prevention of an initial event development into design basis
accidents, a DBA development into beyond design basis accidents through using
the safety systems. It also aims at mitigation of the consequences of the
accident in case of a failure to prevent them by localization of the releasing
radioactive substances.

This level aims at the management of
beyond design basis accidents. The management activities include prevention of
DBA and lower mitigation of the consequences, ensure leak-proof enclosure and
bringing NPP back to a controlled state.


level of activities focuses on preparation of on-site and off-site emergency
action plans and their realization.



Safety system for design basis and beyond design
basis accidents:

In case of development of design basis and beyond
design basis accidents different active and passive safety measures are taken.
The following active and passive safety systems are implemented in V-392M

§  System of the
primary circuit emergency and planned cooldown and spent fuel pond cooling is designed, in particular, for residual heat removal from the
fuel in reactor to the system of the reactor compartment consumers
(intermediate circuit) under all the design basis conditions of the Unit
operation, maintenance of the required coolant inventory in the reactor in case
of a large-break LOCA, emergency primary make-up in the conditions of a
small-break LOCA (D nom 25-80) and medium supply to the spay system;

§  low pressure
emergency injection system is
designed for boric acid solution supply to the reactor coolant system in case
of loss-of-coolant accidents including the break of RCS with a maximum D nom
850 when the pressure in the system goes below the working parameters of the
given low pressure emergency injection system;

§  emergency core
cooling system,
passive part is designed for boric acid solution
supply with a concentration not less than 16 g/kg at primary pressure below 5.9
MPa in the amount sufficient for reactor core cooling before the low-pressure
part of the system of emergency and planned cooldown and spent fuel pond
cooling actuate in design-basis loss-of-coolant accidents;

§  passive core
flooding system is designed to provide the boric acid
solution supply with a concentration of 16 g/dm3 into the reactor core at
primary pressure decrease to 1.5 MPa and less in order to replenish the water
inventory in the core to a safe level, thus providing its reliable cooling
together with the passive heat removal system in loss of coolant accidents in
reactor coolant system including those coinciding with loss of AC power sources
for a period of 24 hours and more. T he boric acid solution supply in the tanks
of passive core flooding system is used to fill up the reactor core barrel and
reactor internals inspection well for the period of refueling;

§  emergency boron
injection system is designed for boric acid injection
into the pressurizer in case of a primary-to-secondary leak to reduce the
primary pressure and create the required concentration of boric acid in the
primary coolant under a BDBA without scram;

§  emergency gas
removal system is designed to remove the steam-gas
mixture out of the RP primary side (reactor, PRZ and SG collectors) and reduce
the primary pressure in order to mitigate the consequences at design basis- and
beyond design basis accidents;

§  primary
overpressure protection system is
designed to protect the RP equipment and pipelines from the gauge pressure on
the primary side under the design basis conditions of Category 2 – 4 and
beyond-design basis accidents due to the operation of the PRZ pilot-operated
relief valves installed on the line for steam discharge out of the PRZ steam
space into the relief tank;

§  secondary
overpressure protection system is
designed to protect the RP equipment and pipelines from the gauge pressure on
the secondary side under the design basis conditions of Category 2 – 4 and
beyond-design basis accidents due to the operation of the SG pilot-operated
relief valves installed on the steamline sections between the steam generators
as far as the shut-off electric motor-operated gate valves, considering the
advance actuation of BRU-A and reactor trip system;

§  passive heat
removal system is designed for long-time residual heat
removal from the core at a beyond design basis accident – loss of all the
sources of AC power supply on condition of the primary and secondary side
integrity retention. Besides, under certain scenarios of beyond design basis
accidents with a primary or secondary leak combined with a simultaneous loss of
all the sources of AC power supply, the PHRS contributes to providing the
required coolant inventory in the primary circuit by the steam condensation in
the SG tubing, the steam being generated in the reactor due to heat removal
from the SG on the secondary side and to the PHRS and the condensed primary
coolant is returned into the reactor.

§  steam generator
emergency cooldown system performs
the safety functions to remove residual heat from the core and cool the reactor
plant down via the secondary side;

§  main steamline
isolation system is designed for quick and reliable steam
generator isolation from a leaky section. T he system is designed for work
under all the accidents that require a SG isolation:

At pipeline breaks downstream of the SGs
as far as the turbine stop valves in the pipeline sections that either can be
isolated and or cannot be isolated from the SG;

At feedwater pipeline breaks downstream
of the SGs as far as the check valves;

At primary-to-secondary leak;

§  double-envelope
containment and
core catcher is designed to retain the radioactive
substances and ionizing radiation within the limits envisaged in the design.


Safety under seismic impacts:

The seismic input parameters are assumed
in accordance with the values provided below:

§  SSE with a frequency once every 10000 years magnitude 8 to
MSK-64 scale;

§  Operating basis earthquake with a
frequency once every 100 years magnitude 7 to MSK-64 scale.

A system of
seismic monitoring and signaling is envisaged in the AES-2006 design that
provides command generation for automatic reactor trip in case of a seismic
input on the ground that corresponds to OBE.

Probabilistic risk assessment:

The results of the probabilistic safety
analysis of Level I confirm meeting the main engineering principles of the
up-to-date concept of defense-in-depth including the principles of functional
and design diversity, protection from common-cause failures, proof-ness against
the operator’s error, physical separation and assurance of a high reliability.

Emergency measures:

Emergency measures include the development of Emergency Action
Plans. The Emergency Action Plans to protect the personnel and the population
are the main regulatory documents to carry out the protective, organizational,
engineering, preventive health-protection and other measures to be taken in
case of an accident in order to protect the personnel & the population,
localize the accident and cope with it. The following steps are taken in the
Emergency Action Plans:

§  Information of the personnel and the population;

§  Commitment of control authorities to action;

§  Radiological and general reconnaissance;

§  Radiological protection;

§  Medical protection;

§  Physical protection;

§  Public order protection;

§  Evacuation activities.




The PSS is a part of a set of
engineering and organizational measures to provide nuclear and radiological
safety at

AES-2006 operation. The task of PSS
includes the following:

§  prevention of unauthorized actions;

§  detection of unauthorized invasion of an intruder into the
secure areas, buildings, rooms and structures;

§  objective confirmation of the information obtained from the
discovering facilities using video monitors;

§  call of response group by the alarm-calling signals from the
guard posts and from the secure rooms, buildings


§  detain (slow-down) of intruder’s advance;

§  suppression of unauthorized actions;

§  monitoring, recording and assessment of the operators’ and
first-line group actions;

§  automated monitoring of the people’s access to secure areas,
buildings and rooms;

§  automated reporting of the staff location;

§  remote round-the-clock T V monitoring of the situation in the
secure areas, buildings and rooms;

§  documenting of the event by video recording;

§  on-line broadcasting of voice data through wire- and radio
communication operative channels;

§  detention of people involved in preparation or implementation of
unauthorized actions.




The system contains the independent
power supply sources, distribution and commutation devices. The emergency power
supply system is a system to supply power to the Unit safety system consumers
under all the conditions of operation including those of loss of working and
backup sources supplied by the power grid.



In this paper, possible architectural approach
towards safety of Rooppur Nuclear Power Plant is dealt. The design of the plant
is developed on the basis of requirements of safety rules and standards of
BAERA, IAEA and other international organizations. The construction of RNPP
introduced Bangladesh to the elite club of nuclear countries. The safe operation of the plant will
increase the country’s economic development and increase our GDP as well. 



The author would like to thank Dr. M. A. Rashid Sarkar,
Professor of Department of Mechanical Engineering, Bangladesh University of
Engineering & Technology for inspiration us and giving us guideline.




 Advanced Reactors Information System
(ARIS) https://aris.iaea.org/PDF/VVER-1200(V-491).pdf

Advanced Reactors Information System (ARIS) https://aris.iaea.org/PDF/VVER-1200(V-392M).pdf

Asmolov, V. G. (10 September
2009). “Development
of the NPP Designs Based on the VVER Technology” (PDF). Rosatom. Retrieved 9 August 2012.

Nuclear Sa


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