VVER Nuclear Reactors

VVER
(Vodo-Vodyanoi Energetichesky Reactor)
CORSO DI LAUREA MAGISTRALE
IN
INGEGNERIA ENERGETICA
Prof. F. Giannetti
A.A. 2017/2018
S. Miani
1

Brief History of VVER
VVER (Vodo-Vodyanoi Energetichesky Reaktor) is basically a pressurized water reactor, where the water
works both as a coolant and as a moderator; in fact, it’s also referred to as WWER (Water – cooled Water
– moderated Power Reactor).
The first VVER were built before 1970 (in the former URSS), but the research hasn’t stopped since then.
These type of reactors have always been updated, thanks to research and experience from previous
ones. The designs start out from the I Generation, and proceed all the way to the III+, which is the current
one (the one in which the EPR and the AP1000 reactors belong to).
The first model of VVER-440 (V230), with a 440 MW electric power, had the most common design which
consisted in a 6 loops layout, each with a horizontal steam generator. Of course, the number of loops
depends on the layout, as we shall see later on.
Basic features (e.g. horizontal steam generators) remain unchanged throughout the time, while others
(like safety measures) have been deeply modified, taking into account the ‘’accidents of the century’’.
For example, the enhanced version of the VVER-440, the VVER-1000 (which has been developed after
1975), incorporates an automatic safety control, passive safety systems and containment systems
associated with western third generation nuclear reactors; therefore, this was not a simple repowering, as
additional emergency devices have been added, consequently improving the safety of the reactor.
The specific plant taken into account is the "Belene" plant (Bulgaria): plant with a V-466B reactor,
evolution of the plant AES-92. In fact, in the V-466B the technical and economic aspects were improved
to meet the requirements of the generation 3+.
The VVER-1200 is the version proposed for future construction, being an evolution of the VVER-1000
with an increase in power of about 1200 MWe and more passive safety functions.
2

VVER in the World
NPPs with VVER reactors have their largest settlements in Russia, but have been built throughout
the world thanks to a 50 years experience Russians have gained over time. The table below shows
the major spots, considering NPPs decommissioned, in function and in construction.
Generation Name Model Country Power plants
I VVER V-365 Russia Novovoronezh 2 (decommissioned)
I VVER V-210 Russia Novovoronezh 1 (decommissioned)
II VVER-440 V-213 Ukraine Rovno 1-2
II VVER-440 V-213 Hungary Paks 1-4
II VVER-440 V-179 Russia Novovoronezh 3-4
II VVER-440 V-213 Finland Loviisa 1-2
II VVER-440 V-230 Bulgaria Kozloduy 1-4 (decommissioned)
II VVER-440 V-213 Russia Kola 3-4
II VVER-440 V-230 Russia Kola 1-2
II VVER-440 V-230 East Germany Greifswald 1-4 (decommissioned)
II VVER-440 V-213 Czech Republic Dukovany 1-4
II VVER-440 V-213 Slovakia Bohunice II 1-2
II VVER-440 V-213 Slovakia Mochovce 1-2
II VVER-440 V-213 Slovakia Mochovce 3-4 (under construction)
II VVER-440 V-230 Slovakia Bohunice I 1-2 (decommissioned)
II VVER-440 V-270 Armenia Armenia-1 (decommissioned)
II VVER-440 V-270 Armenia Armenia-2
3

VVER in the World
While the former table showed the first reactors that were built (basically I and II Generation), the
following represents the ones that belong to the III and III+ Generation. We can see how China,
India and Iran are now present.
III VVER-1000 V-428 China Tianwan 1-2
III VVER-1000 V-428 China Tianwan 3-4 (under construction)
III VVER-1000 V-320 Czech Republic Temelin 1-2
III VVER-1000 V-338 Ukraine South Ukraine 2
III VVER-1000 V-320 Ukraine Rovno 3-4
III VVER-1000 V-320 Ukraine Zaporozhe 1-6
III VVER-1000 V-320 Ukraine Khmelnitski 1-2
III VVER-1000 V-187 Russia Novovoronezh 5
III VVER-1000 V-412 India Kudankulam 1
III VVER-1000 V-412 India Kudankulam 2 (under construction)
III VVER-1000 V-320 Bulgaria Kozloduy 5-6
III VVER-1000 V-338 Russia Kalinin 1-2
III VVER-1000 V-466 Iran Bushehr 1
III VVER-1000 V-320 Russia Balakovo 1-4
III VVER-1000 V-320 Russia Kalinin 3-4
III VVER-1000 V-320 Russia Rostov 1-2
III VVER-1000 V-320 Russia Rostov 3-4 (under construction)
III+ VVER-1200 V-392M Russia Novovoronezh II 1-2 (under construction)
III+ VVER-1200 V-491 Belarus Belarus 1 (under construction)
III+ VVER-1200 V-491 Russia Baltic 1-2 (under construction)
III+ VVER-1200 V-491 Russia Leningrad II 1-2 (under construction)
4

VVER Reactor Types
As stated previously, VVERs have been updated since their first construction, which consisted in
only two units, not built on a serial basis. Since then, a total of 67 reactors have been constructed.
These can be divided into four major groups (the last group is still being developed), even though
each one of them has several different layouts.
• The VVER – 440 were the first units constructed on large scale, both in Russia and in some
European Union countries (Slovakia, Bulgaria and Hungary for example). These reactors served
as basis for the future projects. Their safety features followed the General Design Criteria (US
AEC, 1971), making them similar to the II generation PWRs (in terms of safety principles). All the
other projects followed.
• After that came the VVER – 1000, considered a milestone in terms of innovations. Most of the
operating units belong to this group. We have three different types (V – 320, V – 428, V – 412 and
466), the last two were designed on the basis of the first one. Except for some small changes
related to the energy production system, the rest belong to the safety one, which combines both
active and passive systems. The one that we shall analyze here belongs to this group, it was
supposed to rise in the Belene site (Bulgaria), but its construction was suspended. It was a V-466B
reactor, evolution of the plant AES-92 (V – 412 and 466). In fact, in the V-466B, technical and
economical aspects were improved in order to meet the requirements of the generation III+.
• The VVER – 1200 are the latest evolution, being some units under construction. Most of the
innovations are safety related. Moreover this reactors meet both (other than Russian Regulatory
Documents of course) IAEA requirements and EUR (European Utility Requirements). Concerning
EUR, this means that they can be in fact built in European Union Countries.
5

VVER Reactor Types
• While the VVER – 1200 are the last version being constructed, the VVER – 1500 are the latest in
terms of development. Just like the previous one this design meets Russian Regulatory Documents
(since it will be constructed primarily in Russia) and both IAEA requirements and EUR. Increase of
thermal power, core height and decrease of core heat rate (when compared to VVER – 1000) are
only some of the new aspects of this units. However we will look up at its features later on.
The table below shows the difference, in terms of energy production systems, of the different kind of
reactors.
Type
Thermal
Capacity
[MWth]
Coolant Inlet
Temperature
[°C]
Coolant Outlet
Temperature
[°C]
Average Core
Power Density
[kW/l]
Plant Efficiency,
net [%]
Discharge
burnup
[MWd/kKg]
Fuel
Material Coolant
VVER-1000
(V-466B) 3000 291 321 108 33.7 52.8 UO2Light Water
VVER-1200
(V-392M) 3200 298.2 328.9 108.5 33.9 60 UO2Light Water
VVER-1200
(V-491) 3200 298.2 328.9 108.5 33.9 60 UO2Light Water
VVER-1500
(V-448) 4250 298 330 - 35.7 57.2 UO2Light Water
VVER-300
(V-478) 850 295 325 - 35.3 65 UO2Light Water
VVER-600
(V-498) 1600 299 325 - 35 56.6 UO2Light Water
VVER-640
(V-407) 1800 294.3 322.7 64.5 33.3 47.38 UO2Light Water
6

VVER Reactor Types
From the previous table we can see how the fuel type has remained the same in all the reactors,
same for the plant efficiency. The thermal capacity has increased of course (note that the number to
which the reactor refers, labels the electrical power output), during the years. The outlet temperature
of the primary coolant changes from type to type. From a safety point of view for example, the VVER
– 1000 is the most secure since the outlet water temperature is the lowest (temperature value is the
same as the one we have in the AP1000). Imagining, in fact, a fixed functioning pressure, a lower
outlet temperature means that the coolant is further away from the saturation temperature. However,
there also some cons. we know that the thermal power exchanged in the steam generator depends
on he temperature difference (across the reactor’s core and hence across the steam generator), the
water flow and the thermal exchanging coefficient. Imagining the same water flow and coefficient, with
a lower temperature difference we will have a decrease in the thermal power exchanged. Unless, of
course, bigger steam generators are chosen (just like it has happened in the AP1000). The coolant of
the future VVER – 1500 will have for example an outlet temperature of 330 [℃], similar to the one that
we have in the EPR. So these are the main differences of the various VVER designs, regarding the
power production system.
7

VVER – 440 General Features
This has been the first design to be constructed on
a serial basis, and has also represented a solid
base for the future reactors. The number
represents the electrical output of the reactor (i.e.
440MWe corresponding to 1375 MWth), and the
NPP that we will consider is the Paks Nuclear
Power Plant (that was Hungary’s only NPP). The
plant consists of four VVER – 440/213 nuclear
reactors each with six primary loops, which is
different from the usual western PWR’s layout
(common 4 loop reactors). This wasn’t the only
difference though; in fact if we look at the primary
circuit’s layout, the first thing that we notice is the
uncommon (for nuclear reactors at least) steam
generator’s design, these are in fact horizontally
aligned. Fuel assemblies have also a different
layout (hexagonal), and this leads to a different
assembly organization in the reactor’s core. It’s in
fact common, for usual PWR’s fuel assembly, to
have the 17 x 17 layout (the difference will be
more clear when we will talk about the VVER –
1000). Another big dissimilarity lies in the reactor
building, as we can see from the picture below,
there is a wall that physically separates the
reactor’s pressurized vessel from the steam
generators (located all around it).
8

As stated before, major design features have remained unchanged throughout time, except for small
changes. For example steam generators, reactor vessel and fuel assembly designs haven’t
changed. What has gone under a big change are the safety features (and the safety systems related
to them), especially after the three major nuclear accidents. This is why we shall give only a brief
description of the common features (as we will see them better with regard to the VVER – 1000
reactors), while we shall take a deeper look into the safety systems.
The 3D primary circuit’s layout is presented in the picture below.
9

VVER – 440 Primary Circuit
A schematic view of the primary circuit is shown below. Each one of the six steam generators is
connected to the reactor vessel through the usual lines (cold and hot leg). Moreover the primary
coolant pumps are located one in each loop and twelve isolation valves appear in total. The
pressurizer is connected to the circuit through two surge lines instead of one (typical of western
PWRs), both attached to one loop (being the primary circuit’s components hydraulically connected,
there is no need of having more than one pressurizer). The cold leg of the sixth loop is connected to
the pressurizer’s spray system through a set of valves, its function will be described later on.
10

VVER – 440 Reactor Vessel
The first thing that we notice is the particular disposition of the
nozzles, located at two different heights (hot leg nozzle on top
of the cold leg one). In fact, in western PWRs the nozzles are
placed next to each other; even though sometimes there is a
small difference in height, in order to facilitate natural
circulation, this happens for example in the AP1000 and
MARS reactors. The shape derives from transportation
issues, since it allows easier carriage by trail. We also have
no bottom penetrations, however this is common in all the
VVER reactor vessels.
The core is made up by 276 hexagonal fuel assemblies, with
126 fuel rods each. The control rods are basically composed
by fuel and absorbing material, and they are 37 in total. The
fuel used is 𝑈𝑂2, with a no more than 2.4% enrichment in
𝑈235. The average burn up of the spent fuel is 30 MWd/Kg.
While linear power values are lower than PWRs, this allows
less damage in fuel rods during normal operations and during
AOO. Those are defined as Anticipated Operational
Occurrences, and for no reason they have to be the starting
point of a postulated accident. This is a general requirement
for all the nuclear reactors.
11

VVER – 440 Steam Generators and Main Coolant Pump
These are horizontally shaped components, and it’s one of the biggest differences with respect to
western PWRs. They are placed all around the reactor vessel, behind a physical wall. It‘s basically
an heat exchanging component, where the primary fluid (that has to be cooled down) passes
through 5536 tubes, bent in U – shape coils (in the AP1000 steam generators, the tube number is
approximately 10000, but the power output is way larger). The material used for steam generator’s
tubes is typically Inconel, due to its anti corrosion properties. Such a property is in fact important
due to biphasic fluid outside the tubes. In this case the material used is a particular steel
(08H18N10T steel) which has a protective passive oxide layer, hard to remove, and that makes it
better than AISI304. As usual, dry saturated steam is produced and sent, through the steam header
to the turbine.
The primary fluid is being circulated through vertically shaped centrifugal single - stage canned
pumps. We have the same kind of pumps in the AP1000 reactor, even though in this last case they
are attached, upside down, directly to the steam generator. These components are very tight, in fact
one unit accommodates both the hydraulic and electrical part, making it difficult to have leakages.
This is in fact the main problem for the usual primary pumps used in NPPs. Having very high
pressure differences (between hydraulic and electrical units), the primary fluid (very high pressure
and temperature) tends to leak and flow into the motor compartment. So a whole recirculating
system (also linked to the Chemical and Volume Control System) is connected to these pumps, to
avoid these leaks. The system used to cool down the pumps is the Intermediate Cooling System,
which is comparable, in terms of functions, to the Component Cooling System described into the
PUN project.
12

VVER – 440 Pressurizer
This component is needed when dealing with a pressurized system, since it provides both a
volume in which liquid could flow in due to fluid expansion for example (insurge transient) and an
amount of water that increases the primary system’s capacity and could eventually get out through
the surge line (outsurge transient). Through this component we can control the system’s pressure.
Thanks to the heaters (entered laterally, while they are usually inserted from the bottom), for
example, the pressure could go up (increase of vapor percentage); or could be decreased using a
spray system, connected to one of the cold legs right after the pump’s delivery (in order to have
high pressure liquid). It is usually linked to the hot leg thanks to a particularly narrow pipe, in this
case we have two. In normal operating conditions we have thermodynamic equilibrium between
liquid and vapor phase inside the pressurizer; now, since the temperature difference between the
water in the pressurizer and the one that flows through the hot leg is smaller, if compared to the
one between pressurizer’s water and cold leg’s liquid, the component is linked to the former in
order to stress less the materials. Moreover control systems are placed inside the pressurizer,
connecting it to the cold leg could cause fake measures due to pipelines vibration (induced by the
primary pumps). Last but not least, those control systems help operators understand what is
happening into the pressurized vessel, connecting the component to the cold leg would lead to a
delay in getting those informations. Since changes into the core induce coolant modifications, the
closer this component is to the reactor vessel, the better. One could think, at this point, that the
best option would be to link it directly to the pressurized vessel, but it’s of fundamental importance
that the vessel has as low penetrations as possible, being among the most important components
in a NPP (belongs to the safety class 1). It is hence important that this component has high
capacity, and in fact it does; this is a feature that we also find in the AP1000 and EPR
pressurizers. In case of high pressure, the pressurizer discharges the steam into the Pressure
Relieve Tank through a set of pipes with valves on it. The system is presented in the next slide.
13

VVER – 440 Pressurizer’s Safety Relief Valves
This system is similar to the one of the EPR (Over Pressure Protection), where the steam is also dumped
into the PRT. However, a different solution has been chosen for the AP1000, since the steam ends into the
IRWST (In – Containment Refueling Water Storage Tank) through two spargers. The system below is the
one belonging to the VVER – 440.
14

VVER – 440 Main Isolation Valves
These are wedge type gate valves, used to isolate, so it could be easily removed, a part of the loop for
mantainence. This is why we have two for each one of the loops. An electric motor provides energy to
command the valve. Pressurized water (coming from the make up system) provides the seals necessary to
prevent leakages into the disconnected loop.
VVER – 440 Primary Coolant Pipes
These pipes have an outer diameter of 560 mm and a thickness of 32 mm, their main material is austenitic
steel. Those are the pipes that link together the main components of the primary circuit, the connection is
set up by Argon welding. In this particular welding, called GTAW (Gas Tungsten Arc Welding), a gas is
used as a shield; however it’s very expensive (due to Argon) and needs highly specialist operators.
VVER – 440 Confinement
As we have seen in the first slides, the main confinement is basically a building that holds all the main
(from a safety point of view) components. These are the ones described above. The reactor vessel is
physically separated, through a cylindrical wall, from the rest; and so are the six steam generators. The
volume of this building is around 14000 𝑚3 and can withstand a pressure of 0.1 MPa. The air recirculating
system, equipped with a cooler, cools down the environment.
15

Considering this NPP, the containment represents the third barrier against the release of radioactive
material into the surrounding environment. The second barrier is represented by the physical
boundaries of the primary circuit, while the first is the cladding of the fuel rods. It is though important to
note that nowadays the first barrier is the fuel pellet itself. In fact, when produced by powder sintering, a
certain amount of voids are left on purpose so the first gaseous fission products end up there. However,
when the pellet cracks, the gasses get out and start filling the fuel rod.
Anyway, we have seen how all
the different reactors present a
similar design basis, and a
certain number of new features
that have been updated. This
also works for the containment.
We shall now consider a
specific containment scheme
(VVER – 440/230) and describe
its main characteristics, knowing
that (at least for the 440s) it
won’t change that much. Its
scheme is shown below.
16

In this case the free volume of the containment is equal to 52500 𝑚3, much larger than the one
considered previously. There are two pressure limits, an overpressure (settled to 150 kPa
maximum) and an under pressure limit (set to 20 kPa). The material used for the barrier is
concrete; the inside surface is covered with a steel liner, this is necessary since it has to be
protected from eventual vapor in the atmosphere. Everything is hermetically sealed, in order to
stop any possible leakages. One system worth describing is the BCS (Bubbler Condenser
System) with a series of air traps. The pictures below show the 3D view of both the containment
(right hand side), which is basically the sealed part; and the BCS (left hand side), connected to
the containment thanks to a corridor through which the air – vapor mixture is supposed to pass.
17

This is basically a passive system, and it comes into play when the pressure in the containment
reaches a certain level. The system is activated by means of pressure difference between the sealed
area and the space containing the components of the BCS. What happens is that the air – vapor
mixture (since we will always have a certain amount of air in the containment atmosphere) enters the
BC well first and then the tanks. Here, in contact with water, it condenses thus reducing the pressure.
1. BC Shaft, 2. BC units with water siphoned seals, 3. level of water solution (boric acid), 4. Dual check
valve, 5. Air trap, 6. Lockable check valve
The atmosphere of these tanks is made up
by the part of the mixture that doesn’t
condensate, this flows through a dual check
valve (following the rise of the pressure) into
the air trap and it remains there. This system
can be considered as the analogous of the
Ice Condensers used in some NPPs, but also
similar to the Pressure Relief Tank connected
to pressurizers.
18

VVER – 440 Safety Systems
High Pressure Safety Injection System
The main system used to compensate losses from the primary circuit is the Make Up System, through
make up tanks (with a volume of 18 𝑚3
), which is responsible for minor leakages. When those become
larger, the HPSIS comes into play, providing the necessary volume of water. It also helps lowering
reactivity worth of the system, to maintain the reactor subcritical. During normal operating conditions this
system is in a standby state.
The system supplies boric acid solution,
through a set of six pumps divided into two
groups that work separately because
connected to different parts of the primary
circuit. The injection starts when either the
pressure in the primary circuit drops (below
9 Mpa) or when the pressure or the water
level drop into the pressurizer (below
respectively 11.77 MPa or 700 mm). When
this happens, one pump per group passes
from reserve state to operation mode. In
case one should fail, another pump starts
automatically.
19

Confinement Spray System
The goal of this system is to decrease
pressure and temperature inside the
confinement building, for example during
a LOCA. It is made up by a boric acid
solution tank, that provides the liquid
compound used to cool down the
environment’s atmosphere. The fluid is
pumped up through spray nozzle thanks
to spray pumps. This system is also in a
standby mode during normal operating
conditions, but becomes operative when
pressure reaches values close to 15 kPa
or higher. When a LOCA takes place, the
coolant flows out of the primary boundary
and so do all the radioactive elements
contained in it. In order to reduce 𝐼131
presence in the atmosphere, a particular
solution (KOH + 𝑁2 𝐻4 + 𝐻2 𝑂) is sprayed
through the nozzles.
20

This design is considered as an evolutionary update, since
in the rest only small changes (if compared to this) are
applied. While most of the components related to the
power production system remain the same, great changes
are introduced regarding safety systems. Innovative
passive safety systems add up to the ones that come from
previous experiences (comprehending both passive and
active systems). The reactor itself, thanks to these
systems, becomes more reliable with respect to DBA
(Design Basis Accidents) and BDBA (Beyond Design Basis
Accidents). This is in fact a very important accomplishment,
since when it comes to these accidents the requirements
are really strict (especially concerning DBA, since their
frequency is higher than the one of BDBA, still being really
low). New features apply to the main components and
previous active (or non existing) systems become passive.
For example the Passive Heat Removal System (similar to
the one used in the AP1000) or the Passive Quick Boron
Injection System). Another important feature is the so
called ‘Leak Before Break’, also used in the EPR. The main
pipes of the primary circuit are basically constructed so
they start leaking before a complete break down; this gives
the operators a useful amount of time to act. It is mostly
used on the main pipes, since a rupture of those would
cause major problems.
Reactor thermal power [MW] 3000
Number of Loops 4
Primary pressure [MPa] 15,7
Secondary pressure [MPa] 6,27
Inlet coolant temperature [°С] 291
Outlet coolant temperature
[°С] 321
Coolant flow rate [m3/h] 84 800
Average linear heat rate of fuel
rod, W/cm 166
Average burnup (in stationary
fuel cycle) [MWd/Kg] 43
Steam capacity [t/h] 4 x 1470
21

VVER – 1000 (V – 466B)
The features described in the preceding slide are
related to the generic VVER – 1000 design. We shall
now concentrate on the V – 466B version, and
analyze all the components and systems related to it.
This particular reactor was supposed to be built in
Bulgaria (city of Belene) and the design belonged to
the last update of the VVER – 1000 series, the AES –
92. The project has of course been approved by the
Russia Regulatory Authorities (meaning that it has
received the license), but the EUR certificate has also
been given to it (which means that it could be sold
and built in European Union Countries) and, last but
not least, it meets the standards set up by the IAEA.
The primary system’s layout does not differ so much
from the one that belongs to the previous version
(VVER – 440). Connected to the reactor pressurized
vessel are four loops (instead of six), each with a
steam generator (maintains the horizontal design), a
centrifugal pump, a hot let and a cold leg. The
pressurizer is in this case linked to the loop number
four through the surge line. The system’s electrical
power output is equal to 1060.00 MWe
22

VVER – 1000 Primary Pipes
As usual, the hot leg (with a length of 10 m) is the one that connects the reactor vessel with the steam
generator; while the cold one (26 m in length) connects the outlet nozzle of the steam generator to the
suction of the primary pump, and the discharge of the same pump to the inlet nozzle of the pressurized
vessel. The internal diameter of these pipes has been chosen in order to reduce pressure losses,
considering the existing flow rate (21500 𝑚3/ℎ, per loop), and it has been set equal to 850 mm. the
thickness of the walls of those pipes is 70 mm. The pressurizer is connected to the hot leg of the fourth loop
through a 426x40 mm pipe; while its spray injection line to the cold leg of the third loop through a 219x20
mm line.
The main material used for the pipes is a high alloy
high temperature steel called 10GN2MFA
(10ГН2МФА), mainly composed by Carbon (0.08 –
0.12 in % for grade), Silicon (0.17 – 0.37), Manganese
(0.8 – 1.1), Nickel (1.8 – 2.3) and Molybdenum (0.4 –
0.7).
For the internal surface a layer of corrosion resistant
steel is used (04Х20Н10Г2Б). Iron is the basis
element, together with Manganese (1.8 – 2.2),
Chromium (18.5 – 20.5) and Nickel (9 – 10.5).
23

VVER – 1000 Reactor’s Core
As in the previous design, the fuel element (Fuel Assembly) has an hexagonal geometry. Each element
holds 312 fuel rods with an average enrichment of 4.45 % wt. (percentage in weight) of 𝑈235. The grade
of enrichment depends on the position of the fuel assembly. The single element also contains a fixed
number of empty positions (even though when empty those are filled with water) that house control rods.
These are attached to the Rod Control Cluster Assembly (just like in western PWRs). Each one of these
RCCA has 18 control rods made up by Boron Carbide (𝐵4 𝐶) and Dysprosium Titanate (𝐷𝑦2 𝑂3 𝑇𝑖𝑂2), both
absorbing materials. The latter, when located in the lower part of the control rod, enhances the life of the
component.
The average burn up (considering fresh fuel,
since it changes during the operation) is equal
to 52.8 MWd/Kg of fuel, higher than the one
considered before (regarding general VVER –
1000features).
Once fresh fuel is inserted into the reactor’s
core, 325 (those are effective days) days go by
before refueling operations are needed.
On the right hand side of
the slide we can see the
cross section of a single
fuel assembly. The design
is of course hexagonal and
both fuel rods (in blue) and
control rods (in purple) are
shown.
24

VVER – 1000 Fuel
The main fuel used is, as usual, 𝑈𝑂2 but it is adaptable to MOX, meaning that this too
could be used (just like in AP1000 and EPR). In general a time interval of 1 year in
needed for the refueling, while the standard cycle for the fuel is four times bigger.
We have seen how fuel assemblies and the reactor’s core have different layouts,
between VVERs and western PWRs. The Russian reactor’s fuel assemblies have an
hexagonal geometry and so does the reactor core. It can be easily seen from the picture
below.
25

VVER – 1000 Fuel
An important feature of VVER – 1000’s fuel assemblies is the presence of thermocouples positioned on top
of the FA. These components give useful informations about the outlet (out of the core) temperature of the
primary fluid. It is in fact fundamental to keep this temperature below the saturation one (related to the
functioning pressure) and possibly uniform (throughout the radial dimension).
The measured quantity
changes of course with
the type of reactor. It is
the same for all PWRs,
being the saturation
temperature the limit. In
BWRs for example, the
quantity considered is
the vapor quality.
26

VVER – 1000 Fuel
We have seen how the fuel assembly layout of VVERs is somehow different
from the one that belongs to PWRs; of course the former hasn’t always been
like that. Just like the other components it has been updated (new design is
shown on the left hand side of the slide), for example by adding
thermocouples. The distance grid (picture below) is such that it decreases
difference temperatures along the radial axis and it ‘’sets’’ higher boiling
margins (which means that we are operating under more safe conditions).
Moreover the fuel assembly can be eventually repaired if damaged in some
way.
27

VVER – 1000 Fuel
Moreover, the new features have been included, regarding
the fuel:
• The number of control rods has been increased (passing
from 61 to 121);
• The steel, as a fuel assembly material, has been replaced
by Zircaloy (widely used in western PWRs) with the
addition of zirconium corners (to avoid the twisting of the
assembly along the vertical axis);
• A better compromise for the water – fuel ratio has been
found;
• Use of ne refueling scheme (IN – IN – OUT type);
• Burnable poisons in the core have been replaced by a
uranium – gadolinium mixture. Tis may for example help
reducing the initial concentration of boron in the coolant;
• A debris filter (right hand side picture) has been added to
the fuel assembly.
28

Western PWR Fuel
In this case the fuel
assembly’s layout is
different (we have a
17 x 17 configuration
per FA), and so is the
one of the reactor’s
core. This is shown in
the picture on the
right hand side of the
slide.
The left hand side picture shows a typical
PWR’s FA, with the spider (in yellow)
previously described, that holds the fuel
rods that will be putted in a certain FA.
The horizontal plates are the grids
through which fuel and control rods pass-
The fuel rods are the ones that occupy
most of the space (in red on the boarder
in the picture, but also all round the
control rods).
29

VVER – 1000 Operating Conditions
The NPPs are normally associated with base load of 100% of nominal power, but this doesn’t mean that
they cannot be used in different operating conditions. The present plant could be also used in the load
follow mode. The table below shows different working conditions.
Conditions Notes
Steady state conditions that comprehend
±1% 𝑁 𝑛𝑜𝑚 . These changes can take place at
rates of 1%𝑁 𝑛𝑜𝑚/s.
Power variations and then back to the initial value.
It is possible to operate with two or three reactor
coolant pump sets (note that there are four in
total).
Allowed power variations not larger than
±5%𝑁 𝑛𝑜𝑚 at the same rate as the preceding one.
Power variations at rates smaller than
5%𝑁 𝑛𝑜𝑚/min, considering changes not bigger than
±10%𝑁 𝑛𝑜𝑚.
The variation rate is kept under 5%𝑁 𝑛𝑜𝑚/min, but
with power deviations that go from 20%𝑁 𝑛𝑜𝑚 to
100% of 𝑁 𝑛𝑜𝑚.
30

VVER – 1000 Main Components
Reactor Vessel
This component has been designed on the basis of the VVER –
440. The general layout has remained unchanged. We mainly
have Control Rods (1), with their relative guide tubes; the reactor
cover (2) that cannot be welded to the rest of the structure, since
the vessel has to be opened during refueling operations. Then we
have the reactor chassis (3), on the outside. The inlet and outlet
nozzles (4) which have the same disposition as the previous
designs, one on top of the other. The reactor vessel (5), located
inside the chassis. Finally we come across the active reactor zone
(6) where the fuel rods (7) are. As for all the PWRs, the control
rods are inserted from the top, this is why we find that hat – like
structure on top of reactor cover. The layout, just like in the VVER
– 440, makes the component easier to transport both by rail or by
sea. This is a great advantage since we are talking about a
structure that is approximately 19 m high (comprehensive of the
hat) and 4.5 m wide. Moreover, the insides are settled so it would
be easy to make inspections. Being one of the most important
components, from a safety point of view, internal check ups are of
fundamental importance, especially when welds are present (we’ll
see later that the vessel is made up by cylindrical ‘’pieces’’ welded
together).
31

The picture on the right hand side of
the slide shows a common feature in
pressurized water reactors. As we
know, the inlet water usually flows
outside the barrel, downward towards
the lower plenum; here the water,
coming from the inlet nozzles, gathers
and flows upward across the core
(inside the barrel now) and out through
the outlet nozzles. Here we basically
have the same thing. The inlet water is
colored in green, while the outlet water
is colored in pink. We can see how the
two nozzles are physically separated
one from the other. Moreover the
reactor core is separated, all around,
from the inlet water by the barrel.
32

The ones presented in the picture below are the five main structural components of the reactor vessel. As
we have said before, the vessel it’s not forged as a whole, instead different shells are forged separately
and then welded together. For example the shells that hold the nozzles (four for each shell) do not come
with the rest of the structure. By nozzles we mean the openings that connect the inside of the vessel to the
primary pipelines (cold and hot legs). Most of the pieces are then tighten up with the use of ring sealing
gaskets. The main material (15X2HMΦA) is of course a high temperature resistant steel, similar to the one
used in the previous version. While the internals are made up by corrosion resistant steel. The rest of the
units have the same function as in normal PWRs; for example the core baffle is used as a displacer and as
a sort of shield, located between the core and the core barrel. It is though important to notice that here, the
perforated structure (on the bottom) is only the barrel, being the vessel integer. In normal PWRs we have
holes through which in – core instrumentation passes. The situation is even more critical in BWRs, since
the control rod’s guide tubes also passes from the bottom.
From left to right we have: reactor vessel, upper unit, protective tube unit, core barrel and core baffle.
33

Primary Coolant Pump
In this case we have a typical reactor coolant pump
(GCNA – 1391), in fact it’s a vertical centrifugal one
stage pump. Those are usually one stage pumps, but we
can also have more. A flywheel is located on top of the
pump (connected to the pump’s engine). This
component is important during electrical blackouts,
without it the pump would almost immediately shut down
before the intervention of the emergency diesels. The
flywheel’s provides inertia so the pump continues to work
for a definite amount of time (depending on the
flywheel’s material and dimensions). An anti – reverse
mechanism is also use, to prevent the rotor from going
backwards (and the liquid with it). There is also a sort of
heat exchanging system, to cool down components such
as the engine, and a lubrication system for the bearings.
In both cases water is usually used, to prevent fire
accidents (caused by oil). A thermal shield is always
used in these pumps, since the fluid elaborated by them
has high temperature.
1. Electric Motor, 2. Laminated Coupling Torsion Bar, 3. Internals, 4. Top Spacer, 5. Bottom Spacer,
6. Support, 7. Pump Casing, 8. Flywheel
34

Pressurizer
The design of this component is similar in almost all the PWRs,
only the heaters disposition changes, being in this case
horizontally inserted (power input of 2520 kW). The pressure is
15.7 Mpa. What changes the most is the full volume (capacity
equal to 79 𝑚3 ), becoming usually larger with successive
projects. This allows to threat important transient more safely
(having a a greater amount of water that could be eventually
used, for example); the water inventory is 53 𝑚3. Carbon steel
has been used as the main material, with a layer of corrosion
resistant austenitic steel that has to ‘’deal’’ with the liquid and
vapor phases. The hot leg is connected to the lower nozzle of the
pressurizer through the surge line. We have three nozzles on the
top part. Two are linked to the spray system (to lower the
pressure), and are independent from each other. The upper most
nozzle opens to the OPP (Over Pressure Protection) system,
same as in the EPR (where vapor is discharged in the Pressure
Relieve Tank) and in the AP1000 (where vapor is discharged in
the IRWST), using three relief valves.
1. Surge Bottle, 2. Neck, 3. Internals, 4. Vessel, 5. Tubular
Electric Heater Unit, 6. Nozzle, 7. Support
35

Steam Generator
This is the component (PGV – 1000MK type) that differs the most from the ones used in western PWRs,
as described in the case of the VVER – 440. The layout is pretty much the same, except for some
features, since the power output of the reactor is in this case larger. The capacity of the steam generator
has been increased, in order to have a larger water inventory. The steam pressure is 6.27 Mpa, and the
system has a capacity (related to the steam) of 1470 t/h. The steam moisture never reaches values over
0.20 %. The number of tubes, in which the primary coolant passes through, is doubled with respect to the
VVER – 440 (10978 in total), being this number closer to the one of the EPR’s and AP1000’s steam
generators. The U – tubes (U – coil shaped) are 16 x 1.5 mm in diameter and are made up by austenitic
steel, to protect them from corrosion. The tubes (heat exchanging part) are completely submerged. Some
internal components are attached using the argon – arc welding.
36

The component has a certain number of manholes, to let operators
in to check internals. For example, in vertical steam generator one
of the manholes is located under the tube sheet. This component is
in fact critical from the safety point of view, having a large number
of holes that pass through it (holes that hold the tubes). Just like
the reactor vessel, the steam generator is also composed by a
different number of shells, forged together to form the ‘’vessel’’;
while all the nozzles are simply welded to the structure.
The table below presents the water characteristics.
1. Vessel, 2. Heat Exchanging
Surface, 3. Primary Side
Collectors, 4. Main Feedwater
Distribution Devices,
5. Emergency Feedwater
Distribution Devices, 6. Steam
Distribution Perforated Plate,
7. Submerged Perforated Plate
Primary water inlet temperature [°C] 321
Primary water outlet temperature [°C] 291
Feedwater temperature [°C] 220
The temperature difference of the primary water, through the
steam generator, can be assumed as equal to the one present
when the fluid crosses the core (even though pressure losses
and pump pressure need to be also considered). The water
inlet temperature is similar to the one that the water has (on the
average) in the AP1000 (right out of the core), this leads to
safer operating conditions.
37

Let’s now take a closer look to the particular steam generators used in VVER reactors. We shall proceed
on analyzing the pros and cos of this layout, with a final table where characteristic of both horizontally (for
VVER – 440 and VVER – 1000) and vertically shaped steam generators are presented.
The following description regards mainly steam generators from VVER – 440, though the solutions found
apply mostly to VVER – 1000; moreover considerations about newer steam generators shall be presented
at the end.
With respect to vertical layout
steam generators, these present
no denting, no fouling and no
cracks due to corrosion caused
by primary water. However,
corrosion problems exist but
without them, these components
(regarding U – tubes) could
work for up to 35 years. As we
have seen in the first slides, the
power plant (nuclear island) is
very particular, this makes it
difficult, for these components, to
be substituted. It is a problem
that cannot be underrated, since
it limits greatly the working life of
the plant.
38

Anyway, with respect to the vertical steam generators, the corrosion of the external
surface of the tubes is less violent since in this case no boiling crisis happens (this is due
to the horizontal disposition of the pipes). The vibrations are less strong, since the
velocity of the fluid is moderated and it is easier to remove the sludge from under the
tube bundle due to increase spacing under the tubes. There are also some major
problems, that in some cases cannot be avoided. The cold collector has some corrosion
and cracking issues, related to the manufacturing process. This component represents
the link between the cold leg and the inside of the steam generator, through a series of
holes (perforations along the tube) to which the tubes are attached. These attachments
are referred to as ‘ligaments’, the problem is that they tend to undergo a deformation,
during the lifetime of the steam generator, and this leads to asymmetric stress of the
collector and subsequent deformation. This was a major issue for VVER–1000.
Both VVER – 440 and VVER – 1000 have serious corrosion cracking problems of the
tubes. These are caused by elements and compounds found in the water that crosses
the steam generator outside the tubes. The ones that cause most problems are iron
corrosion products and copper compounds, these are found both in the feedwater and in
the condensate system. The corrosion products are also found in the primary liquid and
their concentration can be used in order to know if some fuel rods have been damaged.
In fact, if their concentration is higher than the one of radioactive products, it means that
(more or less) no rods have leakages; otherwise it means that a certain number of fuel
rods have been damaged.
39

Using particular materials makes it possible to avoid the transportation of these
elements/compounds. For example corrosion resistant materials can be used for the
piping in order to reduce iron transportation. For the copper compounds, either
stainless steel or titanium has to be used for the condenser tubing system.
The corrosion cracking issues could be also avoided thanks to controls done on the
water chemistry. On the primary side for an instance, the radioactivity concentration in
the liquid could be reduced. Moreover, the quantity of corrosion products in the liquid
could also be reduced on the second side.
Concerning VVER – 1000, since the issues exposed above cannot be reduced to
zero, a different strategy had to be considered. This implies innovative methodology
regarding replacement of steam generators (when it is not possible, anymore, to keep
them in operating conditions) . These methods comprehend:
• Primary pipes have to be clamped before being cut;
• Components such as nozzles and the terminal part of pipes have to undergo a
precise machining process;
• Reliable technology regarding the welding process;
• Prediction of the radiation dose.
40

The table below presents the main features of the VVER (horizontal design, both for the 440 and the
1000 reactors) and the PWR (vertical design) steam generators.
VVER – 400 VVER – 1000 PWR
Type Horizontal Horizontal Vertical
Heat Transfer
Coefficient [𝒌𝑾/𝒎 𝟐 𝑲]
4.7 6.1 6.7 – 8.5
Recirculation Number 4 – 6 1.5 – 1.9 3 – 6
Barrier between
primary and
secondary circuit
One Step One Step Two Steps
Heat Exchanging Tube
Material
08H18N10T 08H18N10T Alloy – 600, 690 or 800
Collector/plate 08H18N10T 10GN2MFA,
08H18N10T
Cladding
Low – Alloyed steel,
tube material cladding
Shell 22K 10GN2MFA Low – Alloyed steel
Supports 08H18N10T 08H18N10T Carbon or stainless
steel
41

VVER – 1000 Auxiliary Systems
Distilled Water System
This system provides treated (in this case that undergoes a distillation process) to particular
components and to the primary coolant. This is common in NPPs since the water has to e pure, in order
to be used. Of course in some cases the water is just recycled in the plant and used again. This is a
common routine especially regarding borated water (primary fluid). This water endures particular
processes, in the liquid radwaste, before being used again.
High Temperature Purification System
In this system a series of filtering bed filters are used to remove corrosion products from the primary
coolant, the capacity is equal to 400 𝑚3/ℎ. The water is treated at high temperature and without
pressure drops; it is not always like this, in fact sometimes the cooling down of water is a must. It
happens for example in the CVCS, but we shall discuss this later.
The products that have to be removed are activated corrosion products, like Co – 58 and Co – 60 (with
an half life of 70.78 d and 5.27 years respectively). The first one usually comes from Nickel based
alloys, while the second one from impurities of metallic materials. This element is generated from
neutronic activation of iron (fuel rod’s cladding).
42

Low Temperature Purification System
This system (with a capacity up to 60 t/h) treats blow down primary water (it is basically water that has
been removed, on purpose, to avoid concentration of dangerous products) and leakage water. This
system takes care of the boric acid remaining contained in the water, at the end of the operating time.
Boron Concentrate System
This system provides boron concentrate (with a concentration of 40 𝑔/𝑑𝑚3) to the primary water,
through the Chemical and Volume Control System. We have in fact seen (related to western PWRs
like the PUN) how the CVCS has several links with the BRS (Boron Recycle System), in order to
restore the right boron (boric acid) concentration in the coolant. In the CVCS the water may cross a
certain number of demineralizers with exchanging resins (there’s a three way valve, so the fluid might
not pass through the resins). The last demineralizer acts on the boric acid concentration, absorbing
part of the acid if necessary. A boron concentration measurer is located on one of the pipes, to
measure boric acid quantity in the water; if needed the acid will be eventually added to re establish the
right amount.
System for Supply of Reagents to the Primary Coolant
This system exchanges (gives and takes) chemical reagents with the Chemical and Volume Control
System to maintain precise chemical characteristics of the primary coolant.
43

Chemical and Volume Control System
Considering both identified (like the ones that come from the sealing of the reactor coolant pumps) and
unidentified leaks, this system provides an amount of water such as to compensate the lost one. This
water is supplied to the primary system. However, its main goal is also the control of the chemistry of
the water. Typically there is only one CVCS, that takes and provides treated water to all the loops; the
connections are mixed so the water gets uniformly distributed to the system. It is generally withdrawn
from the cold leg of one loop (it has to be cooled down, so it is more convenient to take it from the
‘’cold’’ source) and supplied to another; though a supply link also exists between the source and the
reactor coolant pump (where a part of the water is sent). The water has to be cool down and the
pressure lowered. Of course the temperature is first taken down, otherwise we would probably have
vapor formation (if we were to lower the pressure and then the temperature we could have saturated
water), then the pressure and again temperature. These processes are needed if we wish to use
demineralizer resins, since those are made up by organic materials and cannot be crossed by high
temperature water. Then the water is sent to the VCT (Volume Control Tank), where radioactive
gasses (noble gases) are removed thanks to a spray injection system and other elements are added
up (like hydrogen to control oxygen concentration, and nitrogen for the VCT atmosphere). At this point
the water may undergo other treatments while it is sent back to the primary system.
44

VVER – 1000 Containment
We have already seen how, regarding the VVER – 440 reactor, the containment represents the third
barrier against the release of radioactive material. It usually consists of two barriers divided by an
annular space, where ventilation (natural or forced) is guaranteed. In this part we shall proceed on
analyzing the containment designed for the VVER – 1000 and mark the main differences with the one
that belongs to the AP1000s, EPRs and to the RBMKs. The picture below* shows, graphically, the
designs belonging to different generation VVERs, regarding the reactor containment building.
*Couldn’t find the source of the picture.
45

Considering the VVER – 1000, we have the usual layout; two barriers (first and second wall)
separated by an annular space. The first wall (the one that actually contains the primary
system) is a cylinder that ends with a dome (hemisphere shaped). Both walls are made up by
concrete, but with differences. In the first one we have pre stressed reinforced concrete, with a
carbon steel liner. The liner is the one that ‘’sees’’ the inside of the containment. This is the
structure that, in case of an accidents, comes in contact with vapor and radioactive materials.
The space between the two barriers is 2.2 meters wide. This space is considered necessary.
We may have a ventilation system, in case of a forced one. Sometimes radioactivity measures
are taken in the annulus, to get informations about what is happening inside the containment.
And in this case, of pre stressed concrete, this space is needed for the control of the pre
stressing system. The second wall (the one used to protect the primary system and all the
components related to it) is a reinforced concrete cylindrical structure that terminates with an
hemisphere (the dome, just like the first wall). The following slide shows the design of these
components.
46

On the left hand side the
crosscut of the containment
building is shown, with the
main components of the
primary system:
1. Horizontal steam
generator;
2. Reactor coolant pump;
3. Containment building;
4. Refueling crane;
5. Control rod assemblies;
6. Reactor vessel.
47

While the first wall’s main job is to keep isolated the primary system and related components, the external
wall protects what is inside it from external events. These events may be natural (strong wind and
earthquakes for example) or artificially caused (explosions, airplanes crashes, etc..). The picture below
describes what has just been said. Note that the containment belongs to the VVER – 1200.
48

VVER – 1000 Containment Building
The picture below shows the complete reactor containment building, as seen from the outside.
49

VVER – 1000 Containment Comparison: AP1000
Regarding the AP1000, we can consider the basic layout similar to the previous one. The first barrier
can be divided into two parts, being the fuel (ceramic form) the first one and the fuel rod’s cladding
(ZIRLO) the second one. The boundaries of the primary system represent the second barrier and the
containment vessel (not the reactor vessel) the last one. This barrier is made up by steel and contains
the primary circuit. This wall is separated from the outer barrier by an annular space. Air enters this
compartment through holes and flows inside it thanks to natural circulation. The steel barrier is cooled
down by water, dropped from the Passive Containment Coolant Water Storage Tank situated on top
of the third barrier. This helps condensate the vapor that could eventually be released in the
containment’s atmosphere; the water is then reintegrated into the tank again. The picture below
shows the plant layout and in particular the location of the PCS gravity storage tank.
The open penetrations (in the steel
barrier) have been reduced by 50% in
number; moreover this wall can be
isolated if necessary. The ‘isolation
issue’ has been a source of discussion
(regarding safety requirements in the
EUR) with respect to the water tanks
located on top of the first wall, since it
does not ensure an hermetically sealed
confinement.
50

VVER – 1000 Containment Comparison: AP1000
The picture on the left
shows how the decay
heat (produced in the
reactor) is transferred
and removed. We can
easily see the steel
vessel (being this the
third barrier) and the
outer barrier with the
PCS gravity drain tank.
The white arrows show
the direction of the air
flow inside the annular
space.
51

VVER – 1000 Containment Comparison: EPR
Same sequence for the EPR. We have the fuel pellet first and then the cladding; only this time the
latter is made differently (M5), to reduce losses from fuel rods. The second barrier is again
represented by the primary boundaries. The main components (especially pipes) composing the
primary system, have been designed considered the so called LBB (Leak Before Break), in order to
avoid direct rupture of the pipeline (especially with regards to double end guillotine breaks). In fact,
this way these components tend to leak before a major break; this gives the opportunity to notice
the issues and intervene. the finally we have the third barrier, shown on the right side of the slide.
52

VVER – 1000 Containment Comparison: EPR
The last barrier is, as usual, the reactor building. The structure is composed by a first wall (pre –
stressed concrete) lined with steel. The steel is basically a gigantic vessel that encloses the primary
system and related components. Then we find the annulus, whose atmosphere is kept below the
atmospheric pressure. This way losses from the main containment will end up in it and will remain
there, since air from the outside will tend too to get in. Finally we have the external wall, made up by
reinforced concrete, to protect everything inside it from external harms. The plant layout is shown
below. the pool (IRWST) on the bottom is inside the reactor building).
53

VVER – 1000 Containment Comparison: RBMK
In this case the situation is quite different because these reactors did not have a proper
external containment. In fact, a characteristic of these reactors was the accomplishment of the
refueling operation with the reactor in power mode (also done by the CANDU reactors). This
could be done thanks to cranes that were located above the reactor. In order to keep all these
components inside the same building, the structure was raised (nearly 70 meters tall) making it
difficult (and expensive) to protect it with an external containment. So basically, excluding the
usual barriers, we have an hermetically sealed steel vessel that contains most (some pipes are
located outside the confinement) of the primary systems. This containment is filled up by
nitrogen, which keeps the high temperature graphite from interacting with the oxygen in the air.
The moderator is also a sort of barrier, since it shields out the radiations coming out of the
core. All this is kept inside a protection structure (made up by reinforced concrete).
On the left, a close up of the reactor vault
is shown; while the complete reactor
building is presented in the next slide.
Here, as numbered in the picture, we
have:
1. Top cover, removable floor of the
central hall; 2. top metal structure filled
with serpentine; 3. concrete vault; 4. Sand
cylinder; 5. Annular water tank; 6. Graphite
stack; 7. Reactor vessel; 8. Bottom metal
structure; 9. Reactor support plates; 10.
Steel blocks; 11. Roller supports.
54

VVER – 1000 Containment Comparison: RBMK
1. Graphite stack;
2. Metal structure ‘S’;
3. Metal Structure ‘OR’;
4. Metal structure ‘E’;
5. Metal structure ‘KZh’;
6. Metal structure ‘L’;
7. Metal structure ‘D’;
8. Drum Separator;
9. MCP bowl;
10. MCP electric motor;
11. MCP discharge
valve;
12. Suction header;
13. Pressure header;
14. distribution group
header;
15. Lower water
pipelines;
16. Steam/water
pipelines; 17. Down
comers; 18. Refueling
machine; 19. Crane in
central hall;
20. Pressure
suppression pool.
55

VVER – 1000 Safety Approach
In this part we shall analyze the safety systems adopted in the VVER – 1000’s design, and then
compare them with the ones used in the AP1000 and EPR.
We have previously seen how the VVER – 1000 reactor is the most wide spread, among Russian
reactors. This means that the requirements that it has to satisfy are not only those imposed by the
Russian federation (since the majority has been built in Russia), but also established by different
organizations. For example the standard imposed by the IAEA or the EUR (in order to be built in
Europe). The safety concepts are based on both active and passive systems. The design has
been intended to be much more simple and reliable, considering for example the manufacturing
and construction of components. This is basically the first big difference among the three reactor
types.
In the AP1000 for example, the approach towards safety is mostly of the ‘passive type’. This
doesn’t mean that it does not have active safety systems; in fact these systems appear (we
basically have the ones that a second generation reactor has), but they are not safety related
which means that they only exist in order to improve the safety requirements. So for example, the
NRC requirements are met only with passive safety systems.
For the EPR the situation is completely different since the reactor is based on active safety
systems. We shall see how, in order to reduce the probability of undergoing an accident, certain
technical solutions have been adopted with regard towards particular components or circuits. The
redundancy concept is what characterizes the most these reactors.
56

VVER – 1000 Defense in Depth
The term defense in depth refers to the act of interposing a certain number of barriers
(physical ones) that, in case of accident, will prevent or reduce the release of radioactive
material in the surroundings, thus protecting people and the environment. Of course we are
not only talking about the NPP’s personnel, but also civilians. There are, related to these
barriers, engineering systems design to provide their correct functioning and protection.
Those barriers we are talking about, are the ones described in the ‘Containment’ part. The
barrier is composed by the fuel matrix (porosity that allows a certain amount of fission
gasses) and the fuel rod cladding. The second protection is done by the primary coolant
boundary, and the third is basically the containment (two walls divided by an annular
space).
57

VVER – 1000 Defense in Depth
In the VVER – 1000’s design, five levels of D. in D. were considered:
1. Level 1 is based on the prevention of the AOO (Anticipated Operational Occurrences).
These are basically events that are related to normal operating conditions, and it is
important that they do not degenerate into a postulated accident (this has to be avoided
in any reactor type).
2. Level 2, where the prevention of the DBA (Design Basis Accidents) is done by normal
operation systems.
3. Level 3 in which the prevention of the DBA passes to the safety systems.
4. Level 4 where the dealing of BDBA (Beyond Design Basis Accidents) is considered.
5. Level 5 based on the emergency planning.
As the level increases, the condition that the plant has to face becomes more severe. We
can understand this by considering what are defined as ‘power plant conditions’. It’s a series
of situations related to the normal/accidental status of the plant during its operation. They
are presented in the following slide.
58

Power Plant Conditions
The updated power plant conditions have been defined by the ANSI 18.2, and are six in number.
Except for small changes, they are basically the same for EUR, IAEA and American regulations.
1. Refers to the normal operating conditions;
2. Comprehends failures and abnormal operations;
3. More severe failures that may happen once in the plant’s lifetime;
4. Design Basis Accidents;
5. Beyond Design Basis Accidents (they normally refer to complex sequences that end up in
accidents more severe than the DBAs);
6. Severe Accidents.
For example, damaged fuel rods (in limited number) fall into the third condition and below; the first two
do not comprehend the loss of integrity of these components. Of course the regulations are usually
very strict with respect to ‘possible’ accidents, while they become less severe as the probability of
happening goes down (BDBA and Severe Accidents for example). In the latter case, a demonstration
of how the plant would manage these accidents is enough.
59

VVER – 1000 Safety Indices
we shall now compare the probabilities of care damage and great releases of radioactive
material, with respect to VVER – 1000, AP1000 and EPR.
 VVER – 1000:
• Core damage: 3.1*10^-7 / reactor / year;
• Large Radioactive Releases: 1.77*10^-8 / reactor / year;
 AP1000:
• Core damage: 2*10^-7 / reactor / year;
• Large Radioactive Releases: 2*10^-8 / reactor / year;
 EPR:
• Core damage: less than 10^-5 / reactor / year;
• Large Radioactive Releases: less than 10^-6 / reactor / year;
60

Safety systems are here both active and
passive. Through their operation, they
have to be able to operate the scram of
the reactor and taking it towards a
subcritical operating condition. Heat
removal (in emergency conditions) both
from the reactor and from the spent fuel
cooling pool. It is important to notice that,
with regards to the spent fuel pool, an
emergency diesel (to maintain the heat
removal even during a electrical black out)
is required for these components after the
Fukushima Accident.
These systems are associated with
different accidental situations. For
example, passive systems (and systems
installed for the management of BDBA)
are used in case of Beyond Design Basis
Accidents while together active and
passive are considered when dealing with
Design Basis Accidents.
A 4 x 100% redundancy
characterizes the safety systems
(same as in the EPR.
61

Emergency Core Cooling System
This system (passive part, shown in the picture)
provides the cooling down of the reactor when, in
case of LOCA (Loss Of Coolant Accident), water is
lost from the primary system. It comes into play
with accumulators, when the pressure goes below
5.9 MPa (remember that we are still talking about
pressurized water reactors), supplying water
mixed with boric acid (in order to make the reactor
subcritical). The main components are of course
the accumulators (four in number), valves and
pipelines that connect the tanks to the primary
circuit. The accumulators are usually pressurized
(tanks with water and nitrogen atmosphere), since
they have to inject water into the primary system.
The volume of the tank is around to 60 𝑚3, with 50
𝑚3
of borated water and 10 𝑚3
of nitrogen gas.
The accumulators are among the first components
that intervene when a cooling down and a supply
of water is needed; in fact pipelines are usually not
that long and they are located e the reactor
containment. The primary system together with the
four hydro accumulators is shown in the next slide.
62

Emergency Core
Cooling System
It is important to notice
how in this case the
pipelines (in blue) are
connected directly to the
reactor pressurized vessel.
It is an uncommon design
since the number of holes
into the vessel tends to be
reduced to the main ones.
For example in the PUN
(second generation PWR)
the pipelines end up into
the cold leg, in order to
provide a direct flow into
the vessel.
63

The complete layout of the ECCS is here
shown, together with its components. Usually
check (non return) valves are located along
the pipes, to avoid the primary coolant from
reaching the accumulators. Motorized valves
are also present, being closed in normal
conditions. Usually connections, together
with valves, exist between the connection
pipelines. This may help transfer water from
one tank to another in case of failure of a
valve or break of a pipe.
64

High Pressure Injection System
We have two of these systems, the difference lies on the tank from which they take water. One
contains highly borated water, while the second one provides water with a lower concentration of
boric acid. The components that belong to these systems are pipelines, valves and high pressure
injection pumps. Together with the accumulators and the Low Pressure Injection System, it
composes the Emergency Core Cooling System. It is made up by four lines, each with a pump. The
‘high pressure’ is related to the primary system’s pressure, meaning that the HPIS starts to inject
water into the cold leg when the system is still pressurized (around 12 MPa). Each channel is
connected to one of the loops. This is important because in case of a LOCA one of the loops might
be unavailable and the relative HPIS line will end up injecting water that will probably be lost. The
high pressure pumps most of the time provide water with high pressure but at low rates. In fact,
when this system starts injecting water the vessel is still pressurized, so the head provided by the
pump needs to be large. The flow, on the contrary, might be small since the loss of coolant can’t be
large if the primary system is still at high pressure.
65

Low Pressure Injection System
This system is either connected to one of the legs (usually the one linked to the pressurizer) or
to the accumulator’s channels. The structure is similar to the one of the High Pressure Injection
System, but with some technical changes. The water is withdrawn from a specific tank or from
a pool located under the reactor vessel. In the PUN for example, the LPIS used to get water
both from the RWST (Refueling Water Storage Tank) and from bilges located on the floor. The
pipes that connect the bilge to the pump were supposed to have a reduced number of valves
on it, and are instead protected all around. In fact the water taken from the floor will probably be
in thermal equilibrium with the vapor inside the containment, thus more valves will cause bigger
pressure losses, and we might have problems related to the NPSH of these pumps.
The LPIS pumps are, in opposition with the ones that belong to the HPIS, characterized by
small head values (around 2 MPa and 5 times smaller than the one provided by the high
pressure pumps) since when this system starts working, the pressure in the vessel has reached
low levels, and large flows (nearly 10 times bigger than the others) to compensate big losses in
case of a LOCA for example. When the system’s pressure reaches 2 MPa the water gets
injected at a rate equal to 250 𝑚3/ℎ, while it reaches 700 𝑚3/ℎ at 0.98 MPa.
66

Reactor Control and Protection System
This system acts in order to prevent an accident from happening, and eventually limit the damages
in case it cannot be avoided. During normal operating conditions the power level is kept constant
thanks to this system. In case of AOOs, it lowers the reactor’s power bringing it below a safety limit.
When an accident occurs, the SCRAM is also done by this system, meaning that the control rods
positions are controlled by it.
Quick Boron Injection System
This system belongs to the ones used for Beyond Design Basis Accidents, in particular when the
reactor trip system fails. It is used to take the reactor in a subcritical state without the SCRAM (since
it’s not good for the reactor to undergo a lot of SCRAMs). There are basically four tanks (that
contain boric acid), each connected to the primary system and a set of valves. The solution is
injected into the main pipes of the primary system, so the boric acid will flow directly into the reactor
core.
67

Safety Boron Injection
System
The Injection of boric acid by this
system occurs when both the reactor
trip system and the previous one fail
to start. Of course, its job is to make
the reactor subcritical without the
intervention of the control rods. Each
channel of the system is connected to
the HPIS, thus to the cold leg of the
primary circuit. However, the SBIS
has another function, when leaks
between the primary and the
secondary system occur (for example
when one or more pipes of the steam
generators are damaged), it pumps a
water – boric acid solution into the
pressurizer in order to lower the
pressure.
68

Emergency Gas Removal System
It is used to remove gas – steam mixture from the pressurized reactor vessel (upper part), steam
generators and pressurizer, when needed. In fact, it is intended primarily for Beyond Design Basis
Accidents, but it is not forbidden to use it in normal operating conditions or in case of Design Basis
Accidents. The removed gases end up into the relief tank, located inside the containment building (since
it usually contains radioactive elements).
69

Core Passive Flooding System
It provides a boron – water solution to inject into the main circuit in case of primary leakages,
especially when an electrical black out occurs (also considering emergency diesel’s failures). Its
main job is to make the reactor core subcritical and to remove the decay heat. Four groups, of 2
accumulators each, provide the right amount of water for no less than 24 hours. These
components are second stage accumulators (HA – 2), different from the ones used for the ECCS
(HA – 1). The boric acid concentration into the water is somehow similar to the one of the High
Pressure Injection System.
Isolation System of Main Steam Lines
The purpose of this system is to confine a steam generator in case either the steam or the
feedwater lines undergo a break.
70

Steam Generator Emergency Cooldown and Blowdown System
This system may intervene in a series of accidental situations. For example in case of electrical power
black out, failure (breaks) of the main pipes of the primary circuit, damage of the steam/feedwater lines
and consequent loss of heat removal from the primary system and rupture of one or more pipes of the
steam generator (leakages from the primary to the secondary system). Whenever one or more of
these situations happen, the SG ECABS provides decay heat removal and the cooling down of the
core.
Primary and Secondary Overpressure Protection System
The overall system consists of five relief valves (PORVs, which sands for Pilot Operated Relief
Valves), three of which are connected to the pressurizer (primary system OPP) and the other two to
the main steam lines of the steam generators (secondary system OPP). The valves act independently
one from the other. The OPP system is in general very important not only because of safety features
(it helps reduce rapidly the pressure, especially with regards to the primary circuit), but also because
by decreasing the pressure (up to 1.0 MPa in the primary system, considering Beyond Design Basis
Accidents) all he other safety systems may intervene.
71

Passive Heat Removal System
This system is intended to remove the decay heat from the reactor core, when an accident occurs.
Its operation does not depend on the integrity of the primary circuit, in fact it can provide a heat
sink even when some leaks occur in the system (either primary or secondary). When losses arise
into the primary circuit, the Passive Heat Removal System is coupled with the Emergency Core
Cooling System (in this case the HA – 2 accumulators are used). Of course, for the latter to be
used the pressure of the primary system has to decrease below the one in the accumulators (for
these to start injecting water).
The PHRS is mainly composed by four natural circulation lines, each with two heat exchangers.
The part connected to the steam generators, withdraws steam and takes it through two heat
exchangers, the condensate is then discharged again into the steam generator. The cooling down
of the steam is done by air, which flows into the HEs using ducts made on purpose. The air is
basically atmospheric air, extracted from the outside.
72

The power of the system is approximately 200 MW and the total number of heat exchangers is 18. The
simplified scheme of the PHRS is shown below.
73

VVER – 1000 Core Catcher
The main retention and cooling system is located under the pressurized vessel, as usual in any
PWRs (but also in other types of reactors). The difference is that in the case of VVERs, the reactor
pressurized vessel is located well above the ground level, with respect to usual western PWRs
configurations. The fact that the cavity is located right under the RPV means that it has been
designed to hold not only the melted corium, but also fused parts of the reactor vessel (break
through of the corium). The system is also used to diminish the radioactive and hydrogen releases
into the atmosphere of the containment, make the core subcritical and cooling it down.
74

The picture presents a different layout of the core catcher, even though its function is of course the
same as the previous one. As we have seen, related to a single group (for example the VVERs –
1000), we may have plants that differ in some things, in this case the core catcher’s design.
1. Containment; 2. Reactor; 3. Concrete Cavity; 4. Concrete Cantilever; 5. Device for
Coolant Supply; 6. Device for Coolant Removal; 7. Ring Section Heat Exchanger;
8.Basket; 9.Protective Truss;10.Heat Insulation Panels; 11. Air Cooling Channels; 12.
Heat Insulation; 13. Lower Plate
75

In the picture below (left side) the complete layout is shown, reactor pressurized vessel and core
catcher, while the other two images show the particular design of the system.
76

VVER – 1000 Safety Systems Comparison: AP1000
In the ffollowing slides we shall proceed on analyzing the safety systems of the AP1000 reactors
(dwelling on the most important/characteristic ones). But first we shall describe some general aspects
of the AP1000.
The reactor’s power output (3415 MWth maximum, which turn into nearly 115 MWe) is located
somewhere between the one, a bit smaller, of the 900MWe reactors (three loops) and the one of the
1300/1500 MWe reactors designed by Westinghouse (lasts of the second generation kind). At the
start, the reactor was supposed to have a 600 MWe electrical power output (AP600), but almost at the
end of the project it was discovered to be not economically competitive. Thus, two parallel paths were
followed. The one of the AP1000, with small changes with respect to the AP600’s layout and an
increased 20% on the costs. The other one was the EP1000, with three loops instead of two; since big
changes had to be done to add a new loop to the design, the former was chosen.
As we said before, the reactor relies mostly on passive safety systems (even though active ones are
still present), this helps reduce considerably the costs since the number of components needed is
lower than the one that would be necessary if active safety systems were used (considering safety
related components, we have 35% less pumps, 50% less valves, 85% less control wires, 80% less
pipes).
77

The primary system is composed by
two loops, each with one hot leg and
two cold ones. The primary coolant
pumps (canned type) are located right
after the steam generators (vertical
ones) and inserted upside down, in
order to avoid the link between the two
components. There are four of them
and from each RCP a cold leg starts,
ending into the reactor vessel. The
pressurized reactor vessel is similar to
the usual design of the one of the usual
western PWRs. The material used is
carbonium steel (A105, A106), since it
has better mechanical properties with
respect to stainless steel . However,
the inside surface of the vessel is
covered by a stainless steel liner (AISI
304), this part is the one that interacts
with the water. The primary circuit’s
layout is shown on the right hand side
picture.
78

Passive Core Cooling System
This system (PXS in short) is basically responsible for the residual heat removal, water solution (with
boric acid) injection and depressurization of the primary circuit. The water inventory needed for the
cool down of the reactor core (other than taking it to a sub critical state) is provided by three different
components (passive). The CMTs (Core Makeup Tanks), usual accumulators and the IRWST (In –
containment Refueling Water Storage Tank). The CMTs are full of borated water, and their only aim
is to increase the water amount of the primary system, in case of accidents. The IRWST is a big pool
located on top of the reactor vessel. These components are connected directly to the pressurized
vessel so that, even if one (or more) of the main pipelines breaks, the core will still receive water. The
water will flow into the vessel by gravity, but since the IRWST is at atmospheric pressure, the one in
the PRV has to decrease below a certain level (around 0.18 MPa). The depressurization is provided
by the ADS (four stage Automatic Depressurization System). The ‘stage’ is related to the fact that
these valves open at different pressure levels and are mounted on pipes of different diameters.
These valves are similar to the pressure relief valves connected to the pressurizer, there are though
some differences. In case of the ADS for example, the pipes end up in the IRWST instead of the
Pressure Relief Tank. Moreover, three of the four stages are connected directly to the head of the
pressurizer, while the last one on the hot leg.
79

Passive Core Cooling System
The overall PXS system is shown on the right. The top red lines are the ones that connect the pressurizer
to the IRWST (through the ADS), while the bottom one is the natural circulation line responsible for
removing part of the heat from the primary system. This part will be treated in the following slides.
80

Passive Residual Heat Removal
This system belongs, just like the other components described previously, to the PXS and its purpose
is to remove (passively) the decay heat after the reactor shutdown (especially in this case). The
system is shown in the picture below, though the pipe that enters the IRWST is actually connected to
a heat exchanger located inside the tank.
The first thing we notice is the total
absence of pumps (hence the
‘Passive’) thought out the circuit.
The water flows thanks to natural
circulation, from the hot leg (since
the goal is to remove heat, it is
better to do it with water at higher
temperature) to the heat exchanger
and then back to the primary circuit
(through the steam generator). The
HE is located inside the IRWST,
this allows to have a great amount
of water for the heat removal. We
can also notice how the cold and
hot legs are located at different
heights (not like in the VVER –
1000), in order to enhance the
natural circulation.
81

The complete system is shown in the picture below. We can see the three stages of the ADS that
connect the pressurizer to the IRWST. The pipes discharge the steam through a set of spargers.
It’s basically horizontally placed tubes (0.5 m in length and 20 cm of diameter) with a series of
holes in it (around 1 cm of diameter). Thanks to these holes the steam enters the tank not in a
violent way, and the condensation is smoother. If the steam had to be dumped into the tank only
through a single hole, water – steam oscillations may have occurred.
In fact, the instant condensation
of steam, at the pipe end, would
have created a depressurization
that could have called back
water from the tank. This water,
proceeding upward would have
ended up in hitting the steam
coming from the pressurizer, to
end up again in the IRWST. And
so on.
This kind of phenomenon
caused problems in a Mark II
containment (BWR).
82

Passive Containment Cooling
This is also a passive system, and its goal is to
cool down the inside of the reactor containment
building so that the pressure doesn’t exceed the
limit value (design pressure settled at 400 kPa).
The cooling down of the building (composed by
steel) takes places thanks to the air that flows
on the outside (between the steel and the
concrete walls), driven by natural circulation.
The air is taken both from the outside, but it’s
also the one that comes from the evaporation of
the water dropped from the above tanks. From
the right hand side picture, we can easily see
how the tanks located above the steel
containment, drop water on the containment,
which removes the heat and cools it down. For
an instance, the steam generated in the IRWST
due to the heat exchange is released into the
containment’s atmosphere and then condensed
(when it comes in contact with the containment’s
wall).
83

In Vessel Retention
The main goal, in case of molten
core, is to keep everything inside
the pressurized vessel. In order to
do this in the AP1000, we have the
IVR system. When a part of the
core starts to fuse (the core will
probably start to melt in a
particular zone, instead of doing it
altogether), it flows downwards. At
one point, it falls on the lower core
support plate. If this component
fails to maintain the corium (note
that the vessel is still full of water),
the melted mixture will start to drip
on the components below the
support plate, ending up on the
secondary core support. The
falling sequence can be seen from
the picture.
84

In Vessel Retention
It is obvious now that the inside cooling will not be enough, it is in fact necessary to start cooling
the vessel from the outside. This is done in a particular way, shown in the picture below.
The cavity between the reactor
vessel (bottom part) and the
concrete basement is flooded
when a plug opens and water
starts to get inside. This water
keeps the vessel refrigerated, by
removing the heat generated by
the corium. The heat removal is
taken care of from the outside.
The steam generated in this cavity
exits from the designated steam
vents (right under the primary
nozzles). It then enters the
containment atmosphere, where it
condenses (containment wall is
permanently cooled down from the
outside) and can be used again.
Only doubts regard the heat
removal mechanism in these
conditions.
85

VVER – 1000 Turbine – Generator System
The turbine (below) is typical for this NPP (K – 1000 – 60/3000), and it’s of course a steam turbine with
steam superheating. It has a speed of 3000 rpm (rotational) and a power of 1000 MW (nominal value).
The total length, turbine plus generator, is 68.8 meters (51.45 m only for the turbine). The steam comes
from four steam lines (we have four steam generators), with a pressure of 6.27 MPa (typical pressure
inside PWR’s steam generators, secondary side) and at a rate of 5880 t/h.
The Generator (three phase) has the usual configuration. The cooling of the windings is done by two
different fluids, water for the stator and hydrogen (which is also used to cool down the stator core) for
the rotor. Electrical parameters are 24 kV of voltage and 1000 MW of active power.
86

VVER – 1000 Spent Fuel Pool
Once the spent fuel is discharged from the reactor’s core, it goes through a series of check ups to
see if some elements have been damaged. If this is not the case, the FAs are placed inside the
spent fuel pool, next to the core barrel. The pool is in fact located inside the containment building
since it contains radioactive material. A different tank is provided for the damaged fuel, after the
elements have been sealed in specific bottles. A polar crane (for nuclear use) is installed on the
upper part of the containment, for the spent fuel transfer. The crane is shown below.
The fuel is then transferred from
the SFP to the Spent Nuclear
Fuel Storehouse (SNFS), for dry
storage. The spent fuel resulting
from a 10 years operation of two
units can be stored in this
facility.
However, the goal is to keep it
inside for all operating life of the
units.
87

VVER – 1000 Radwaste Handling Systems
Liquid
The liquid wastes (radioactive and non) deriving from the plant operation are usually treated into
this system. They are first stored in tanks (mostly materials containing short lived radionuclides)
and then processed so they can be sent to the reprocessing system for solidification (where
cement is usually added). These kind of systems are usually alike in all PWRs. Usually the
liquid wastes are separated based on their chemical and radiological properties. this separation
enables to use a specific treatment for each category. For example, borated radioactive
(primary system for the most part) liquids are processed and recycled so they can be used
again (when possible). Non borated but still radioactive liquid pass through an evaporator and a
filter. The distilled solution goes to the waste (but only after controls on the residual activity)
while the mud deriving from the filters and the concentrate resulting from the evaporator are
sent to the solid radwaste processing system.
Gaseous
This system usually comprehends not only units for the processing of radioactive gasses, but
also components able to treat the hydrogen contained in the fluid that arrives. In western PWRs
for example the gas passes (thanks to a compressor) through a component in which the
hydrogen is recombined (in a controlled way), in order to reduce its concentration. Then the fluid
is stored in decay tanks, to further reduce the activity. Finally, the radioactivity is checked and if
the values are below certain limits, it’s released.
88

VVER – 1000 Solid Radwaste Reprocessing System
The radioactive wastes received by this system are first of all not only solid ones. We have seen
in fact how the liquid radwaste handling system is also linked to this one, of course in this case
wastes are usually cementified before storage. Anyway, in general the wastes processed here do
not derive only from normal operating conditions, but also from accidents or maintenance works.
Just like in the previous system, the treatment depends mostly on the composition and
radioactivity. For example low and medium activity wastes undergo different treatments than high
activity level ones. In particular, the former are transported into barrels and taken into a facility, in
order to be processed and reduced in volume. The latter are transferred to the same facility
(Reprocessing and Storage Building) in special containers.
89

VVER – 1000 Building and Structures
The main buildings of the analyzed NPP (Belene site) are the reactor, turbine and the auxiliary
reactor ones plus emergency electrical supply and safety systems buildings. The site comprehends
two units. The general VVER – 1000 power plant layout is shown in the picture below.
90

The 1200 MWe VVER reactor is the latest design fully completed. We shall give a brief description
of the main features, especially for those that have been improved with respect to the previous
ones.
Just like the VVER – 1000, it’s a four loop each with a pump, an accumulator and an horizontally
shaped steam generator, there is one pressurizer for the entire primary circuit, as usual.
Pressurized reactor vessel, steam generators and pressurizer dimensions have been increased.
This allows to have a greater water inventory available in case of an accident. The thermal power is
3200 MWth (but could be increased up to 3300 MWth). Pressure of the primary (16.2 MPa) and
secondary (7.00 MPa) have been increased, and so have the coolant inlet (298.2 ℃ ) and outlet
(328.9 ℃ ) temperatures. This higher outlet temperature would have caused the system to be less
safe if the pressure would have been unchanged, since it would have been closer to the saturation
temperature related to that pressure. However, the pressure too has been increased, in order to
increase the limit gap. Moreover, higher outlet temperature enhances the exchanged thermal power
between primary and secondary systems. The steam capacity has increased also, passing from 4 x
1470 t/h of the 1000 MWe reactor to 4 x 1602 t/h of the 1200 MWe version, where the factor ‘4’ is
related to the number of steam generators. The number of fuel assemblies (163) and the one of
control rods (121) has remained basically unchanged. Though the average burn up is greater in the
VVER – 1200 (up 70 MWd/kg versus 52.8 MWd/kg), with respect to the VVER – 1000.
91

VVER – 1200 Innovative Features
Reactor Vessel
The design of the pressurized reactor vessel has been improved, extending its life time up to 60
years. For example, diameter and height have been increased in order to reduce the neutron fluence
to the internal walls. With particular arrangements fuel cycle can be extended by 60 days, this is
generally called ‘reactor stretch out’; while the normal fuel cycle can go from 12 to 24 months.
Another important feature is the remote extraction of a single fuel assembly. Even though this was
also possible in the new versions of the VVER – 1000. By doing this, damaged Fas could be easily
removed and repaired.
92

Steam Generators
The layout of the SGs has remained unchanged (horizontally shaped), together with its main
components. However, some of the problems found in the previous versions have been solved thanks
to a certain number of new features. Copper has been completely eliminated from any material, tubes
(of the tube bundle) have been arranged differently and a ethanolamine water treatment have been
applied in order to increase the SG life time (60 years). We have seen in fact how this was one of the
main problems with these power plants, since SG’s working life determined the one of the NPP.
The dimensions have been further
increased, making the primary
system safer. In fact the SG, thanks
to its water inventory, operates in
order to remove the heat from the
core before any other component, in
case of an accident (naturally, in
normal operating conditions that is its
first purpose). Moreover, a series of
sinks have been located inside the
SG to facilitate the elimination of
sludge.
93

Containment
The VVER – 1200 presents a double
containment with a ventilated
annulus between the two walls. This
is a common design for modern
reactors, since it reduces largely the
probability of radioactive releases to
the environment. The picture on the
left shows the pre stressing system
of the building. The inner wall has a
44 m diameter while the outer one
reaches 51.8 m. The pressure that
the system (inside containment) can
withstand is approximately 0.4 MPa
(similar to the one of the AP1000’s
steel containment).
The inner building is such that it can
endure a Design Basis Accident and
at the same time a major earthquake
(for which it was design).
94

The strategy adopted for the VVER – 1000 with regards to safety system’s utilization has been recycled
for the VVER – 1200. this means that both active and passive systems are used when a Design Basis
Accident occurs, while mainly passive and BDBA management systems intervene in case of a Beyond
Design Basis Accident. We shall here analyze the main characteristics of these systems.
Low Pressure Emergency Injection System
This system injects water – boron solution into the primary system in case of LOCA and breaks in the
primary system. The injection starts when the pressure (of the primary circuit) goes below a well
defined value.
Just like the previous layout, the ECCS is composed by three subsystems, two of which are active
while the third one is the passive one. The latter (usually first stage accumulators) provides boric acid
mixed with water (same concentration as in the VVER – 1000s) when the pressure goes down to 5.9
MPa and below. The pressure level is settled up by the nitrogen atmosphere of the tanks in which the
water is kept.
95

Passive Core Flooding System
This system intervenes not only when in case of accidents, but also when the refueling has to
take place. In the latter case, the vessel is filled up with water. When an accident occurs and the
pressure goes below 1.5 MPa, this system starts to inject borated water into the primary circuit.
Especially in case of Loss of Coolant Accident and electrical power black out, the PCFS together
with the Heat Removal System, could keep the core refrigerated for more than 24 hours.
Emergency Boron Injection System
Used for Beyond Design Basis Accidents in order to avoid the SCRAM of the reactor. It’s
connected directly to the pressurizer so that the steam inside it condenses when it interacts with
the water (with boron inside), thus decreasing the pressure.
Emergency Gas Removal System
It works in order to reduce the pressure of the primary system when a Design Basis Accident or a
Beyond Design Basis Accident occurs. The pressure in this case is reduced by removal of gases
(and steam).
96

Primary and Secondary OPP
Both these systems operate in order to reduce the pressure of the primary (POPP) and secondary
(SOPP) systems. The pressure is lowered by a set of valves that open in order to release steam from
the two systems (for rapid pressure decrease). For the primary system PORVs (Pilot Operated Relief
Valves) are located on the pipes that connect the pressurizer to the pressure relief tank (inside the
containment). On the secondary side, valves are mounted on the steam lines in order to discharge
the steam coming from steam generators.
Main Steamline Isolation System
Whenever a leak appears (feedwater line or primary tubes of the tube bundle) this system provides a
reliable isolation from the loss.
97

The System (shown below) has different functions based on the situation. During BDBA and
electrical power black out (primary and secondary systems without any leaks) it removes the heat
from the reactor’s core. However, during a DBA with loss of coolant and power black out, it
provides water to keep the core refrigerated. The water used is basically the steam produced in the
steam generators that, previous condensation, is sent again inside the primary system.
98

SG Emergency Cooldown System
Simply provides cooling and heat removal of the primary circuit (mainly from the reactor's
core) exploiting the secondary side.
99

Core Melt Localization Device
This component provides cooling and storage, for an indefinite amount of time, of the melted
fragments that could result from a core meltdown during a severe accident. The fragments we
are talking about can be a mixture of melted core, in vessel components and vessel structure.
It’s basically a sort of vessel that’s supposed to hold and cool melted material. Once the corium
is gathered in the ‘cavity’, it gets distributed around the volume thanks to non metallic materials
settled on purpose.
100

VVER Nuclear Reactors

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