VVER Nuclear Reactors

VVER (V odo-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 consist ed 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, t he enhanced version of the VVER-440, the VVER-1000 ( which has been developed after 1975 ), incorporate s 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-302 Ukraine South Ukraine 1 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-320 Ukraine South Ukraine 3 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 UO2 Light Water VVER-1200 ( V-392M) 3200 298.2 328.9 108.5 33.9 60 UO2 Light Water VVER-1200 ( V-491) 3200 298.2 328.9 108.5 33.9 60 UO2 Light Water VVER-1500 ( V-448) 4250 298 330 - 35.7 57.2 UO2 Light Water VVER-300 ( V-478) 850 295 325 - 35.3 65 UO2 Light Water VVER-600 ( V-498) 1600 299 325 - 35 56.6 UO2 Light Water VVER-640 ( V-407) 1800 294.3 322.7 64.5 33.3 47.38 UO2 Light 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

VVER – 440 General Features 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 , with a no more than 2.4% enrichment in . The average burn up of the spent fuel is 30 MWd/Kg. W hile 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 t he 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 and can withstand a pressure of 0.1 MPa. The air recirculating system, equipped with a cooler, cools down the environment. 15

VVER – 440 Confinement 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

VVER – 440 Confinement In this case the free volume of the containment is equal to 52500 , 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

VVER – 440 Confinement 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 ), 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

VVER – 440 Safety Systems 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 presence in the atmosphere, a particular solution (KOH + ) is sprayed through the nozzles. 20

VVER – 1000 General Features 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 , 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 (10 ГН2МФА) , 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 . 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 and Dysprosium Titanate ( , 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 usua l, 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% . These changes can take place at rates of 1% /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% at the same rate as the preceding one. Power variations and then back to the initial value . Power variations at rates smaller than 5% /min, considering changes not bigger than ±10% . Power variations and then back to the initial value . The variation rate is kept under 5% /min, but with power deviations that go from 20 % to 100% of . Power variations and then back to the initial value . Conditions Notes 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). Power variations and then back to the initial value . Power variations and then back to the initial value . Power variations and then back to the initial value . 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

VVER – 1000 Main Components 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

VVER – 1000 Main Components 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 (15X2HM Φ A ) 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

VVER – 1000 Main Components 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

VVER – 1000 Main Components 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 ), 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 . 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

VVER – 1000 Main Components 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

VVER – 1000 Main Components 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

VVER – 1000 Main Components 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

VVER – 1000 Main Components 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

VVER – 1000 Main Components 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

VVER – 1000 Main Components 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 [ ] 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 VVER – 400 VVER – 1000 PWR Type Horizontal Horizontal Vertical 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 . 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

VVER – 1000 Auxiliary Systems 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 ) 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

VVER – 1000 Auxiliary Systems 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

VVER – 1000 Containment 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

VVER – 1000 Containment On the left hand side the crosscut of the containment building is shown, with the main components of the primary system: Horizontal steam generator; Reactor coolant pump; Containment building; Refueling crane; Control rod assemblies ; Reactor vessel. 47

VVER – 1000 Containment 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 : 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). Level 2, where the prevention of the DBA (Design Basis Accidents) is done by normal operation systems. Level 3 in which the prevention of the DBA passes to the safety systems. Level 4 where the dealing of BDBA (Beyond Design Basis Accidents) is considered. 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 . Refers to the normal operating conditions; Comprehends failures and abnormal operations; More severe failures that may happen once in the plant’s lifetime; Design Basis Accidents; Beyond Design Basis Accidents (they normally refer to complex sequences that end up in accidents more severe than the DBAs ); 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

VVER – 1000 Safety Systems 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

VVER – 1000 Safety Systems 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 , with 50 of borated water and 10 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

VVER – 1000 Safety Systems 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

VVER – 1000 Safety Systems Emergency Core Cooling System 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

VVER – 1000 Safety Systems 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

VVER – 1000 Safety Systems 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 2 50 , while it reaches 700 at 0.98 MPa. 66

VVER – 1000 Safety Systems 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

VVER – 1000 Safety Systems 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

VVER – 1000 Safety Systems 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

VVER – 1000 Safety Systems 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

VVER – 1000 Safety Systems 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

VVER – 1000 Safety Systems 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

VVER – 1000 Safety Systems Passive Heat Removal System 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

VVER – 1000 Core Catcher 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

VVER – 1000 Core Catcher 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 f following 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

VVER – 1000 Safety Systems Comparison: AP1000 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

VVER – 1000 Safety Systems Comparison: AP1000 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

VVER – 1000 Safety Systems Comparison: AP1000 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

VVER – 1000 Safety Systems Comparison: AP1000 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

VVER – 1000 Safety Systems Comparison: AP1000 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

VVER – 1000 Safety Systems Comparison: AP1000 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

VVER – 1000 Safety Systems Comparison: AP1000 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

VVER – 1000 Safety Systems Comparison: AP1000 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

VVER – 1200 General Features 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

VVER – 1200 Innovative Features 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

VVER – 1200 Innovative Features 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

VVER – 1200 Safety Systems 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. Emergency Core Cooling System 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

VVER – 1200 Safety Systems 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

VVER – 1200 Safety Systems 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

VVER – 1200 Safety Systems Passive Heat Removal System 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

VVER – 1200 Safety Systems 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

VVER – 1200 Safety Systems 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 – 1200 Plant Layout The containment building is shown on the left while the complete plant layout on the bottom. From the left hand side figure we can see in yellow the polar crane (also seen in the VVER – 1000 design) and in red (lower part) the core catcher previously described. 101

VVER Future Perspectives: VVER – 1500 The VVER – 1500 is not completely developed like the previous ones, but when terminated it will have all the features necessary to follow the domestic, European and IAEA requirements. It’s naturally an improvement with respect to the VVER – 440 and VVER – 1000 reactor types, and from these it takes the major characteristics, improved by experience and research. The power output is 4250 MWth, larger than any of the previous ones. While the primary pressure has been decreased with respect to the VVER – 1200, the secondary one is higher. Outlet coolant temperature is higher. which allows to have a greater thermal power exchange in the steam generator (with regards to the temperature difference). We have seen how, design by design, these reactors have been improved. Though most things have remained unchanged, especially considering layouts of the components like reactor vessel, pressurizer, steam generators, etc. Others have changed greatly in order to make the plant more safe, reliable and characterized by a longer service. Component’s dimensions have increased, in order to have a bigger water inventory in the primary system and particular constructing features have been used in order increase component’s reliability (changes in the steam generator’s materials and component’s disposition in order to obtain a longer life time). A great deal of changes have been carried out especially when passing from the VVER – 440 to the VVER – 1000, in fact the latter has been of great innovation in the VVERs industry. In the following slides a table with all the main parameters regarding VVER – 1000, VVER – 1200 and AP1000 is shown. 102

Power Plant Parameters AP1000 VVER – 1000 VVER – 1200 Thermal Power Output [MWth] 3400 3000 3200 Power Plant Output, net [ Mwe ] 1100 1011 1082 Power Plant Efficiency [%] 32 33.7 33.9 Operating Mode Base Load and Load Follow Base Load and Load Follow Base Load and Load Follow Plant Lifetime [ ys ] 60 60 60 Thermodynamic Cycle Rankine Rankine Rankine Primary Coolant Light Water Light Water Light Water Secondary Coolant Light Water Light Water Light Water Moderator Light Water Light Water Light Water 103

Reactor Coolant and Steam Supply System Parameters AP1000 VVER – 1000 VVER – 1200 Primary Coolant Flowrate [kg/s] 14300 23888 23888 Primary Pressure [MPa] 15.513 15.7 16.2 Primary Coolant Inlet T [°C] 279.4 291 298.4 Primary Coolant Outlet T [°C] 324.7 321 328.9 T Difference Across the Core [°C] 45.2 30 30.7 Steam flowrate [kg/s] 1889 5880 1780 Steam pressure [MPa] 5.76 6.27 6.8 Steam T [°C] 272.8 278.5 283.8 Feedwater Flowrate [kg/s] 1889 6000 1780 Feedwater T [°C] 226.7 220 227 104

Reactor Core Parameters AP1000 VVER – 1000 VVER – 1000 Active core height [m] 4.267 3.530 3.75 Core Diameter [m] 3.04 3.16 3.16 Av. Lin. heat rate [KW/m] 18.7 15.73 16.78 Average fuel power density [KW/ KgU ] 40.2 / 36.8 Average core power density [MW/m3] 109.7 108 108.5 Fuel material Sintered sintered UO2 UO2 and UO2 + Gd2O3 UO2 and UO2 + Gd2O3 Cladding material ZIRLO Alloy E-110 Alloy E-110 Outer diameter of fuel rods [mm] 9.5 9.1 9.10 RAs Array 17x17 hexaedral hexahedral Number FAs 157 / 163 Enrichment of reload fuel [% wt.] 4.8 4.45 4.79 Fuel cycle length [Months] 18 18 12 Average discharge burn up of fuel [MWd/Kg] 60 52.8 60 105

Reactor Pressurized Vessel Parameters AP1000 VVER – 1000 VVER – 1200 Inner diameter [mm ] 4038.6 4195 4232 Wall thickness [mm] 203 195 197.5 Design pressure [Mpa] 17.2 17.64 17.64 Design temperature [°C] 343.3 350 350 Base material Carbon Steel Steel 15H2NMFA Steel 15H2NMFA Total height, inside [mm] 12056 11235 3855 106

Steam Generator Parameters AP1000 VVER – 1000 VVER – 1200 Type U - Tube, Vertical Horizontal PGV-1000MPK Number 2 4 4 Total tube outside surface area [m2] 11477 6105 Number of tubes 10025 10978 10978 Tube outside diameter [mm] 17.5 16 16 Tube material Inonel 690 - TT Steel 08 Х18Н10Т У 08H18N10T Transport weight [t] 663.7 330 330 107

Primary Coolant Pumps and Pressurizer Parameters AP1000 VVER – 1000 VVER – 1200 Number of pumps 4 4 4 Pump speed [rpm] 1800 1000 1000 Head at rated conditions [m] 111.3 62.2 Flow at rated conditions [m3/s] 4.97 / 5.97 Total volume [m3] 59.47 79 79 Heating Power of Heaters [kW] 1600 2520 2520 108

Containment Building Parameters AP1000 VVER – 1000 VVER – 1200 Shape Cylindrical Cylindrical and Hemyspherical Dome Cylindrical and Hemyspherical Dome Diameter [m] 39.6 / 44.0 Height [m] 82.3 38.5 42.2 Design pressure [Mpa] 0.5067 0.4 0.5 Design temperature [°C] 148.9 150 150.0 Design leakage rate [V % /day] 0.10 0.3 0.2 109

References https ://www.sciencedirect.com/science/article/pii/S0013468607000333 https://www.springer.com/978-3-319-08547-0 http://www.gidropress.podolsk.ru http://www.splav-kharkov.com/en/e_mat_start.php?name_id=3240 http://www.auremo.biz/materials/stal-sv-04h20n10g2b-ep762.html https://en.wikipedia.org/wiki/VVER http://www.gidropress.podolsk.ru/files/proceedings/seminar6/for/jrc-trampus.pdf https://www.sciencedirect.com/science/article/pii/S0306454914002345 https:// www.iaea.org/NuclearPower/Downloads/Technology/meetings/2011-Oct-Simulators-WS/Tikhonov-WWER-1000.pdf 110

References http ://www.iaea.org/inis/collection/NCLCollectionStore/_Public/32/011/32011699.pdf https://ats-fns.fi/images/files/presentations/2009/ATS_jasentilaisuus_2009_SPAEP.pdf http://www.westinghousenuclear.com/Portals/0/New%20Plants/AP1000/AP1000%20Station%20Blackout.pdf?timestamp=1404842353431 http://www.iaea.org/inis/collection/NCLCollectionStore/_Public/33/011/33011210.pdf http://www.iaea.org/inis/collection/NCLCollectionStore/_Public/28/047/28047194.pdf http://www-ns.iaea.org/downloads/projects/archive/wwer-rbmk-safety.pdf https ://www.iaea.org/INPRO/3rd_Dialogue_Forum/17.Ansari-Pakistan.pdf https://etd.adm.unipi.it/theses/available/etd-11242003-105246/unrestricted/capitolo2.PDF https://www.iaea.org/INPRO/7th_Dialogue_Forum/INPRO_Bykov_3.pdf 111

References http :// www.sar-net.eu/sites/default/files/ERMSAR_2013/Slides/6.3%20Zdarek%20UJV.pdf http:// energypolicy.columbia.edu/sites/default/files/A%20Comparison%20of%20Nuclear%20Technologies%20033017.pdf https:// aris.iaea.org/PDF/AP1000.pdf http:// www.westinghousenuclear.com/Portals/5/Other%20PDFs/Coping%20with%20Station%20Blackout.pdf http://www.nkmnoell.com/industries/nuclear/reactor-building-cranes / https:// www.konecranes.com/industries/nuclear/nuclear-equipment/nuclear-polar-cranes http:// www.rosatom.ru/upload/iblock/0be/0be1220af25741375138ecd1afb18743.pdf https://aris.iaea.org/PDF/VVER-1200(V-491). pdf https:// www.sciencedirect.com/science/article/pii/S245230381730095X 112

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