 OK, zelo smo odvorezno. Zelo smo počkati. Češo vseh prišličaj, izgledaj smo z drugim zelo, kaj je zelo koment. Zelo sem, da bo vse zelo, vsega vsega, zelo izgledaj svačnosti. Zelo sem, da počkaj vsega zelo izgledaj svačnosti. z ga v reaktačnih vsehur等一下ov. Bejel je prezentacije vsehrači Kulan sistem in sošnjati sistem, beto kar je kulan sistem in joveti sljudo backyard systemičnih. imate se mana, skupnja in podjelja reaktorixelaki, in ocetjevanih štih. Teknevne koncepto, ki je tudi sveti VWRivec, kot吹ite, sva sveti je zvanjno izopitela skupnja v SARS-2 vs. Prečo početno, oč predostan, kako je se vs. in SSR-2 vs. for these systems. This is the safety guide that currently we have, and it's called NSG 1.9, was published in 2004, and now is under revision to make the new safety guide consistent with the latest requirement, SSR 2 slash 1, because this safety guide was prepared when still the old requirement NSR 1 were in place. So what are the functions? Let's always start from the function and then going down to the system and components. The function of the actual coolant system, associated system, is first to contain the coolant, because if you don't contain the coolant, you cannot cool. This is a process function, but we have to contain the coolant also because the coolant can be activated and contain some radiactivity. So it is a barrier for the confinement of the fission probe, so this is a very important function. Then we have to remove the heat from the core and from the components in all plant states that we are considered in the design. Then we have to transfer the heat, we remove the core to the ultimated sink and this comes through different states, rather complex chains, not just one heat exchange. And we have to maintain the specified physical and chemical characteristic of the coolant. Of course, when you are talking about physical characteristic, that means the temperature, that means the pressure, and when you are talking about the chemical additive, what should be particularly pure and there are some addition chemical products like the boron accident and so on. So we have to control the chemistry and physics of the coolant. So these are the functions of this complex set of system that we call system associated to the main coolant system. So what are the main components of this SCSAS? First is the reactor coolant system, known as SCS. This is quite known acronym. Then we have the system for the residual heat removal that can be system for normal operation and system for accident. It depends on the design. Some of the system can be partially used for both situations. In other design, they are completely separated. That is really in the agreement with the latest interpretation of the difference in depth. Each plant state should have the dedicated systems and cooling systems. Then we have the chemistry, inventory and reactivity control system. And finally, ultimated sink. When we say ultimated sink, at least in the terminology of the agency, the ultimated sink is the final means. Some means the air, the water. This is the ultimated sink. So it can be either the atmosphere or the water in the ocean or in the river. Normally when we talk about ultimated sink, we mean also the interfacing system with the sea or with the air. So the last part of the cooling chain that really is exchanging it with the water or with the air. So this is the ultimated sink. But there is a definition of the agency. Ultimated sink means water or air. It's not a system. So we can have a variety of drawings showing this cooling system for different kind of reactors. And if you go even on the internet, you can find brochure for many reactors and you can have plenty of these beautiful full of colors, pictures with all the explanations. So I refer very quickly to two or three drawings that we have in the safety guide because they are sufficient to indicate the major systems that we have. So this is the typical for PWR. We can see the reactor cooling system. Then we come back with... I put some colors on the specific piping so we can individuate the system more clearly. Here you can see the reactor pressure lesser than the steam generators. And then the turbine, the condenser including the cycle. And then here we have the emergency cooling systems with a tank. Here we have a passive system. And then here we have two lines. One is for high pressure and low pressure pumps to inject water in different conditions. And then we have the ultimated sink. This is for the residual heat removal system. And this is for the other system. Then here we have the pressure risers with the safety valves. So these are the major typical components of PWR. Of course you understand that in reality the situation is much more complex. If you see real drawings with real piping and valves and pumps that you have in a thing that you can fill a page and it's quite difficult to take some time to understand all the interconnections and all the functions. But this is sufficient for our purpose. So I use this and an equivalent that also we have for the PWR. Also here you can see the pressure, the pressure vessel, the depressurization system, turbine, main cycle and then the ECCS and the ultimated sink and the two different cooling chain. This is just to give an idea of how many systems are in this family. This is another representation for another kind of reactor. This is for a Kandu reactor. This is a little thing probably you are familiar with this concept. There is no typical pressure vessel but the core is located inside horizontal tubes and everything is immersed in a big tank of heavy water that acts as a moderator. And then for the rest is very similar to a PWR. We have a steam generator, pressure riser, primary pump and then the steam goes to the turbine. Here is the coolant system of a typical VVR. This is a Russian design. And that is some, there is a particular feature on this reactor because the steam generators are horizontal. And practically it's the only one that I know that using the solution of the big nuclear power plant. And this is Russian technology while in the western technology all reactors have, all nuclear power plants have vertical steam generators. So there are very different technical solutions and all of them work quite well. So what are the things to consider for the design of all these systems? First, the safety function. They have to perform in different plant state. We never get tired to repeat this fundamental concept. Then what are the postulated initiating events they have to deal with? The safety classification and the associated design and fabrication codes. This is what we talked about this morning. Then the loads and loads combination that can affect the design of these systems. Then the protection against external hazard. Typical example is the seismic. The protection against internal hazard. We mentioned yesterday what these internal hazards are. Then the design criteria. For example, the single failure criteria and others. The environmental conditions in which these equipment have to work in normal states and also during the accidents that we are considering our design and the selection of materials and then requirement for testing, inspection and maintenance. So all these should be clearly in mind to the designer to conceive and design in detail these systems. So now, let's go, limited our discussion to the reactor coolant system. What are the functions of this specific system? First, to transfer the heat from the reactor to the steam generators in the PWR or directly to the turbine in a BWR. Maintain the pressure of the coolant within specified limits. Contain the coolant, providing an effective barrier to the release of reactive material and keeping the integrity of the pressure boundary. So these are the main functions of this system. Reactor coolant system. Here are the major components of the reactor coolant system. And you can see here the least of them. Sorry, because they are very close. The main components are reactor vessel in the case of the PWR, steam generator. The primary side that is very important. Only the primary side is part of the coolant system. It is important because the safety class is different from the tubes that inside the steam generator that part of the primary coolant and the shell of the steam generator that is part of the secondary. So that is something very important also to remember. Then the reactor coolant pump, the pressurizer, the piping, the hot leg, cold leg and surge line, the protection system and the depressurization system. So we will discuss some of these, not all of them, because if we have discussed each of these components in detail, it takes much more time. So you can see here a typical configuration of four loop reactor coolant system. For loop, you see there are four steam generators and four primary coolant pumps. Here you see the vessel and this is the pressurizer connected to one line of the primary coolant circuit. So this is a typical configuration that is common to many, many, many plants. What can change is the number of loops. We have plants with four loops and four steam generators and now we have plants, also big plants with only two steam generators with two loops. So there are different solutions. So this is a typical configuration of the reactor coolant system. So here, coming to our initial drawing, you see the coolant system is this one that is the primary coolant system, the one that is in red. So vessel, part of the steam generator, primary pump, pressurizer and of course the pressurizer is a system itself because a system to control the pressure of the fluid and so on. This is very, very schematic what is represented here and then we have the safety valve to decrease the pressure of the system and then there is a tank where this steam can be condensed. Then we have, as I said before, this is a system to remove the heat, the residual heat when the reactor is shut down, shut down state so we cannot remove through the steam generator and there are here two circuits which is generically called intermediate cooling system and this system normally in different technologies assume different names. Quite common name is for this intermediate is component cooling water system and the second, the later, the system that interfacing with the ultimate sink is very often called essential system water cooling or something similar. Okay, but that is not relevant. What you have to have in mind here there are the chain to exchange the heat is quite complex. It's not just one, so there are several loops that are done on purpose for confining the radioactivity mainly and you understand if you have to duplicate one to make a redundant system of this chain, you have to duplicate everything if you want to make independent. This is done sometimes in good design but not always. In some design they have a common system. The initial part of the chain is duplicated but the final part that goes to the ultimate sink is unique. So it's not really a complete independent installation of the implementation of two complete separated independent systems but there is even worse I would say that some systems that are designed to remove the heat in normal operation in shutdown can be used also as a safety function in the accident so they are safety systems. So as you see this is one case where the independence of level of the defense in depth is really badly implemented because safety system should not have function in normal operation but these are existing reactor many of these are designed in this way but there is one consideration to do about the classification if a system has two function one in accident condition and one in normal operation should be classified according to the most demanding condition so it's classified in design as a safety system even if it's used in normal operation so this should be clear but it's not the best practice to implement our requirement on independence of level of the defense in depth then here in green in green we have the emergency core cooling system that is these are the system to deal with loss of coolant accident so if we have a break in the primary coolant system so we lose the primary coolant so we need to cool to cool the core with independent source of water and independent system so we have normally we have accumulators these are those that act first in very high pressure and they inject very quickly water in the core and they can inject only the water contained in the tank because the tank is normally large tank several is not only one normally there is a series of this and there is a separation between a separation between the water and the gas here the gas is compressed it's pushing the water once we open the valve it's pushing the water when the tank is empty it's empty that's no way the water is finished then we have to rely on system in this case an active system so we have tank we suck the water from and then we have system according to different accidents or different moments of the accident mitigation high pressure or low pressure pump inject the water then here this is the chemical control system and the inventory control system so these curves is I'm sure most of you have seen this curve just to remind you what are the physical conditions of the fluid of the water inside the primary and secondary loop of a PWR or the primary circuit of a BWR here is for example let me come here because otherwise I don't see this is the case of the PWR so we have the primary system here the secondary system you can see that the pressure in the primary system is around 15.2 megapascal that means little more than 150 atmospheres if you are more familiar so this is the operating pressure over PWR and you see the temperature here 288, this is the inlet temperature and this is the temperature outside of the core is around 324 and it can change a little bit this value can be different but it cannot be very much different may I ask you why? why cannot be 350 or 400 the temperature the exit temperature in the PWR sorry? yes, because it wouldn't have water it would have steam it would not be pressurized water but pressurized steam reactor unfortunately in this technology there are these limits there are the properties of the water and we cannot go very up in the temperature that is of course it reflects on the efficiency of the thermal cycle these are basic concepts and then you see here what are the conditions of pressure and temperature in the condenser and for the secondary steam you see we are operating a pressure of steam about 7 MPa, 6.9 and the temperature in this case is 227 and 285 temperature so it is quite low temperature for steam cycle so the efficiency cannot be very high because of the physics but anyway, that is just to give you an idea for what are really the values of the key parameters if we move to the PWR the pressure is not very much different it is 7.3 and this is the range of the temperature and of course we reach the saturation line because we produce the water is boiling and inside the core and we have a steam practically exit of the core or a mixture of steam and water of course is a bi-phase situation so what are the specific requirements as I mentioned yesterday I said in our requirement we have some general requirement that apply to all systems and then in section 6 of SSR 2 slash 1 we have dedicated requirements for each important system we have requirement for the core requirement for the coolant system requirement for the containment and so on so I recall briefly what we say there so the requirement 47 design of reactor coolant system the components of the reactor coolant system for nuclear power plant shall be designed and constructed so the risk or faults due to inadequate quality of materials inadequate design standard insufficient capability for inspection inadequate quality of matter is minimized so practically this says the primary coolant system should be designed very high standard so that reflect should be classified in higher class and then design, manufacture and operate at the high standard level so the pipe work connected to pressure boundary because of course there are many lines many other systems that are connected to the primary coolant system of the reactor coolant system for the nuclear power plant shall be equipped with adequate isolation devices so it is possible to isolate if it is necessary to limit any loss of radioactive fluid primary coolant and to prevent the loss of coolant through interfacing systems the design of the reactor coolant pressure boundary shall be such that flows are very unlikely to be initiated flows in the material in the thickness of the piping or the vessel for example and any flows that are initiated would propagate in a regime of resistance to unstable fracture and to rapid crack propagation thereby permitting the timely detection of flows I will add some information on this when we are talking about the vessel but this is of course maybe many of you some of you are specialist in fracture mechanics so you know exactly what we are talking about this is a quite complex science how the flow propagates inside the material different temperature and how they are influenced by the temperature by the other characteristics of the material of the neutron irradiation and so on but the requirement reminds us that we have to take care of all this situation and we have also to work in a regime and in a situation where if there is a flow flow inside the material this is not propagating of this propagating is not propagating fast maybe it is growing but we can keep under control this is just to remind you about design this is simpler it is very schematic but it is representing the same concept that Tony showed you yesterday in a more complex diagram just to indicate the concepts so we have this can be we have a pressure we want to see we have of course a working pressure operating pressure in the pipe and there is of course there is a pressure that is causing the destruction of the piping so we want to know what is the margin we can rely on so we have there is a value of the parameter that is determined with realistic assumption but also this value for many many reasons can be affected by some uncertainties so we have to to be aware of this if we determine the same parameters but using in our calculation conservative approach we determine a different value and for example this is the value we determine and there are some limits established by the safety authority that add some margin to this or it depends which parameter or they deduct some value which parameters we are dealing with and there is another value for example in case of the pressure the value that can produce a cliff edge effect so can produce the rupture of the component and suddenly generate consequences very serious can ask you, some of you to explain what is a cliff edge effect in general what you understand when we say cliff edge effect because we have to understand the terms that we are using Tony already mentioned mentioned yesterday but can be misinterpreted there is also this concept ok, go ahead please so that means you are in a situation where a small variation of a parameter or more parameter can cause a big difference in the consequences it's like when you are walking this is called cliff edge because it's like when you are walking on cliff in this is a cliff and you are walking you are walking here you are walking here in this direction put like this you are walking here if you stay very close to this cliff and you just lose equilibrium a little bit you fall down if you are here if you are here further than this you lose your equilibrium a little bit nothing happens if you are here if you are here you are even safer because you can move and jump you don't fall so the distance that we have from the cliff is practically the margin or related to the margin in a very intuitive way here the margin is this here the margin is this here the margin is this and this concept can be translated in other parameters that is not distance can be pressure can be and we use this term in our safety standard very often in particular referring to seismic events to external events where we have to include large margin to avoid cliff edge effect that means if the acceleration of your earthquake is a little higher than what you assumed in your design basis the reaction of the plant is good nothing happens so you are protected and the earthquake normally the earthquake is not causing severe cliff edge effect because it is very with the earthquake if you have an earthquake that is growing in intensity you don't fail all your components at once maybe you start one component failed and if the earthquake goes higher another component so on but in case of the flooding the cliff edge effect is very drastic because if you have a barrier until you don't reach the barrier you don't have any leak on the other side but if you increase a little bit then you flood the other side so it's very important so this is something to keep in mind a very intuitive concept but we use this word cliff edge effect extensively in our requirement and for margin I mean on margin there is literature so you can read many many publications in our safety standard we don't make difference between design margins or safety margins we call margins in general we use the same meaning safety margin and design margin so the requirement for the design the design of the reactor coolant system the design of the reactor coolant system shall be such as to ensure that plant states in which components of the reactor coolant pressure boundary could exhibit embrittlement should be should avoid you know what embrittlement is so then we will explain that it is something also this is very important because has very strong consequences on the design and consequences on the life of the plant because embrittlement is really mainly caused by neutron irradiation not only but mainly by neutron irradiation the design of the components contained inside the reactor coolant pressure boundary such as the pumps the pumping pillars the valves path shall be such as to minimize the likelihood of failure and consequential damage to other components of the primary coolant system that are important to safety in all operational state and design basis, accident condition we do allowance made for deterioration that may occur in service practically we say that we have to design high quality and take into consideration all the possible phenomena and the load that we can consider let's go to this I would like to spend just few words I'm sure this is a basic engineering so most of you have seen these things and who has seen these things the variation of the energy that is necessary to break a sample of material in function of the temperature what is called a resilience the resilience is a property of material that is the energy you need to break a sample of this material have you seen this who can come here and explain to the colleagues this curve it's not an exam it's just well if you don't I will try to do something but I mean this this is a phenomena that is very important especially for the vessel because the vessel especially in the area facing the core it's subject to irradiation of high energy neutron fast neutron of course there are shieldings inside there is protection but some of this neutron and they change the property of this material if you see if you consider this is just a typical behavior for steel the upper part but let's see and you see here what is plotted in this in this curve is the energy versus the temperature the energy is the energy that is necessary to break this man of this material of course this is a standardized test it's called a Sharpe test that is theoretically is very very simple is still used but doesn't cover all the situations then I will tell you why what do we measure here we have a pendulum sorry for this we have a pendulum that is at this elevation so as this potential energy we put this specimen here and this is is prepared in a particular way standardized size in the code for the material you find all these characteristics and there is a sort of intake here to facilitate the rupture really is the pendulum and the pendulum breaks the material and reach another elevation that is less than this of course the difference in elevation is proportional to the energy so that is and so we can measure the energy that is necessary to break to break the and we plot this energy this energy we make different test at different temperature and we plot this energy and there is a typical behavior if we stay above this temperature higher temperature we need rather higher energy to break the sample then there is this transition area where the energy is decreasing very fast and then if you could if you operate at the very low lower temperature you can break with the little energy and here we say the material has a ductile behavior here is a fragile or brittle behavior that is very important because if you have a pressurized component and you work you work you work in this area the resistance of the material is much higher if you go here you need very little energy to propagate a flow that is inside the material breaks like a glass that is extreme brittle material of course the steel doesn't behave like this but similarly so this is how the material behaves normally this is the characteristic of the steel is always like this but when you irradiate this material you have a switch of this I should cover this so this curve you see here moves down and the more is the irradiation this curve moves down and moves also in this direction that means in general the energy the energy that you need to break and propagate a flow is much lower than here but the temperature also the temperature where you have this transition maybe you heard the term and need activity temperature so is the temperature in this range that is separating these two areas ductile behavior and brittle behavior so if you look at the values these values here are quite realistic for steel used in the vessel in the vessel so you saw that the vessel start of course with fresh material after so many years so many years of irradiation the temperature the temperature that you have this transition to the brittle approach values above the room temperature so you can really inject like Tony mention this morning if you are not careful you can inject a water at the temperature that brings the vessel in this range and you have the risk for the catastrophic rupture of the vessel because the vessel is pressurized and probably in the vessel there are some flaws inside and this changing property the vessel is one of the main parameters that determine the life of a nuclear power plant because as Tony said the vessel cannot be changed and you have to work also with some margin you cannot risk this you have to be far from this and so this is becoming this is important particularly for the vessel and also for piping all other components so it's something to keep in mind because there is there is always there are some accidents that have been study analyzed because of the injection of cold water in these components Tony this morning mentioned the problem caused by different temperature by delta T but there are also problems in radiation so this is important factor in the primary coolant because don't forget the vessel is part of the primary contains the core but it's part of the primary coolant system and is a component if it fails there is nothing we can do there is a catastrophic consequence so it's something to practically eliminate so the analysis of the behavior of the material is one way to reach the practical elimination of this situation so I spent a little time on this but I think it's worthwhile so here for example I have listed some because I found this listed for a VVR but it is the same for other reactors what has been done in modern reactor to extend the life from 40 years to 60 years dictated of course from the economy of the plant you want to use the plant for longer time so that the cost per year is going down and now the modern plant are designed for 60 years you see all the factors that have been considered practically this is a simplified list probably there are more things probably more things than this so there is a new program placement of irradiative stainless steel directly on the reactor vessel oh I forgot to tell you that during the life the life of the vessel the property of the vessel are tested so we should know at which level of embrittlement we are this is done putting inside the vessel sample of the same material that they have the same irradiation irradiation of the walls of the vessel and periodically when there is refueling maintenance we take this sample out and we make the test with the sharpie and so we know in which situation I think this is now normal practice it's not done everywhere and was not done in the past so we didn't know what the situation was but now I think it's normal practice this is also establishing the code so there are rules very well then there are of course some technological solutions like the limitation of the nickel or the property of materials limitation of harmful impurities in the basic material metals and welds modernizational manufacture technique decrease of neutron flux to the reactor vessel of course to use the flux the fluence in the material you have to put some some shields and of course the very sensitive areas are those areas for example where the weldings are so good technique I think almost everywhere now is to avoid well if the vessel is realized in several in several parts is to avoid the welding in the active zone of the core so should not be welding or welding means also penetrations penetration normally they are also welded so this is of course there are several solutions that are implemented so and then increase of the control reliability of the fluence of the reactor vessel so this is a value this is a realistic value for this so we have an idea of which range of temperature we are really talking about temperature very close to the operating situation so the decrease in NDT so when we switch the temperature when we switch from ductile to brittle behavior of the nozzle zone the nozzle and the penetration zone core barrels to minus 35 degrees centigrade at the beginning of light and plus 50 at the end you see what is the switching temperature is 85 degrees so from the beginning of life to the end of life of the vessel the transition temperature moves 85 degrees so it's a big range so it's something very important so we continue with the requirement requirement 48 the over pressure protection of the reactor coolant pressure boundary every time we have something pressurized we have a system to control the pressure and to relief the pressure of course over pressure protection of the reactor coolant pressure boundary provision shall be made to ensure that the operation of pressure relief devices will protect the pressure boundary of the reactor coolant systems against over pressure and will not lead to the release of radioactive material from the nuclear power plant directly to the environment so every time you have pressurized pressurized system you have to have some valves, safety valves that are necessary in case the pressure increase for any reason and you cannot control of course there are system to control the pressure but if the control system is not capable to control or is not working correctly and the pressure exceeds establish value your relief valve open and discharge the valve discharge the fluid inside the containment either in a tank or in the atmosphere of the containment requirement 49 inventory of the coolant provision shall be made for controlling the inventory temperature and pressure of the reactor coolant to ensure that the specified design limits are not exceeded in any operational state of the nuclear power plant with due account taken of volumetric changes and leakages so also this I think is something very very easy to understand but you see every time you have one of these requirement there are several functions then you have to implement and so your system becomes more and more complex because you keep adding system to system to system so you start for a single pipe with a single pump and the steam generator and keep adding hundreds of other components to preserve the structure integrity of the reactor coolant system boundary by keeping in conjunction with the reactor screm the pressure below the design limit specified for different category postulating event so this is the function we are now in the function of over pressure protection system of course it is a system it has some functions we have to identify this function and then put system to implement successfully this function another function is to ensure the protection of the reactor coolant system structure integrity in case of ATWS of course if you have a transient that requires the screm of the reactor and the trip doesn't work for any reason you understand that the power is still generated so the pressure if you for example lost the secondary side and you don't achieve the proper trip of the reactor the pressure of course is increasing but ATWS is a transient is a transient that is part of our design basis so the plan should be designed to cope with this situation there are different ways to cope but the plan can face this situation either with intrinsic characteristic of the plant or with additional systems to shut down the plant to provide protection against an acceptable load combination of high pressure and low temperature when the reactor coolant system is operated at low temperature protection of RCS equipment with material or less ductility at low temperature so this is the same topic that we have already discussed so what the function of the depressurization system this I want to spend few words on this because is something important and particularly important now that we have introduced the design extension condition in the design so may the function depends on the type of reactor so we distinguish between PWRs and BWRs for PWRs for the practical elimination of the phenomena associated with the high pressure melt ejection in case of severe accident PWRs should be equipped with the fast depressurization of the primary circuit system manually actuated so that means that in addition to the system to control the pressure in addition to the system the safety valve that this charge in case you have an excursion of pressure we want to have the ability to fast depressurize the primary circuit why we want this and when what is the because we do this as extreme the latest measure to deal with the big accident if we have if we have a design basis accident a large loss of coolant, large break in the primary coolant and we for some reason we see the operator can realize that the emergency core cooling system are not working because some problem a loss of power so we know that we are going to a core melt situation the core melt situation cannot be avoided because the core is still producing energy the decay heat we don't remove so what can we do we know that we are going to a severe accident so we choose to depressurize the reactor very fast because this first produce a cooling but second we go to a severe accident that we can cope with because with a severe accident in high pressure conditions we can have some phenomena that we cannot cope with a typical is the direct containment heating if the vessel breaks because of the core melt the core melt in high pressure this material is ejected high pressure in the atmosphere of the containment exchange the energy with the gas in the containment very quickly and can pressurize the containment and explode so this is a phenomena we want to avoid to avoid this the only possibility we have is to depressurize we have a core melt but is in low pressure condition and in low pressure condition we can cope with a rather conventional containment structure so this is a way to eliminate to practically eliminate core melt so the system, the modern reactor have this additional system installed just for this purpose cause is a system for design extension condition but really in a very extreme situation when there is nothing else to do we have all the we detect the symptoms in the plant and we know that we are going to core melt for sure there is nothing we can do is better to depressurize so in this case the operator on the side endopne this valve for example but in low pressure I think this would be a lot of useful for example yes, yes yes, right in low pressure of course if the system is already depressurized doesn't make sense to open again and you are right not precise but also if you think also the heat removal system the residual heat removal system is in some reactor considered part of the ECCS so it is an emergency core cooling system but that means you are in a situation you cannot do anything to control the pressure you don't remove the heat and the pressure is going up up and the pressure doesn't grow if the circuit is open already what is sorry is very low we showed yesterday this probability is in the range 10 minus 6 if you look at the table what was we showed the frequency the frequency of different plant states so this is very low probability that is below 10 minus 6 but who knows exactly when you are going to this range if they say so I mean nothing is impossible unfortunately even to have four molten core at the same time and so it's so this feature now is already implemented in some design of course the probability that you reach this situation is different in different plants because this is the result of all the system that you have in the plant if you have very high redundancy in the heat removal chains they are completely separated with completely independent source of power using diverse technologies I mean you can exclude this you can say I don't have to address this situation because my plant takes care of this but it depends on the design of course then for the you see this is a safety feature the safety feature this is important to remember is a safety feature this is automatic switch after sometimes it is not used I cannot start again as I said this is now supposed to be on is red light push is getting weaker and weaker so that is the point so you see here the safety feature designed to practically eliminate direct containment heating for DCA direct containment heating is this one for BWRs the BWRs are already equipped with a system the current BWR are equipped with a system for fast depressurization normally it is called ADS automatic depressurization system because they use the depressurization system to deal with some design basis accidents in order to allow the operation of the low pressure system there to depressurize the reactor otherwise the low pressure system so you see in the BWR practically the same system allow the operation of the low pressure coolant injection function is a safety system designed for mitigation of design basis accident so in the BWR is designed for this very rare, very ultimate design extension condition here is designed to operate in normal in design basis accident so they have to kept in mind these differences ok this is depressurization system what it looks like for some reactor safety valve and so on so requirement for the design of the reactor coolant system requirement for the clean up requirement 50 adequate facility shall be provided nuclear power plant for the removal from the reactor coolant of radioactive substances including activated corrosion products and fission products deriving from the fuel and not radioactive substances cause you are familiar with this during the operation you produce some crates you remove some material and this material is activated and for some reason this material can accumulate in some areas and create high radiation level area to take this in consideration and you have constantly to remove this material and clean the water the capability of the necessary plant system shall be based on the specified design limit on permissible leakage of the fuel with a conservative margin to ensure that the plant can be operated with a level of circuit activity that is as slow as recently practical and to ensure that the requirement are met for the active releases to be as slow as recently achievable and below the authorized limits on this charges this is linked to what we said yesterday for radiation protection of the operator so, we have seen this already I don't know why this is here this is here because we talk about this cleaning up system as shown here in the left in the left side the requirement 51 removal of residual heat from the reactor core another important system means shall be provided for the removal of residual heat from the reactor core in the shutdown states of a nuclear power plant such the design limits of the fuel reactor coolant pressure boundary safety are not exceeded so I have some example here the system we are talking about now is this is sorry is this a yellow one and of course in the PWR there are different approaches now I want to go into details this because this is very design specific in the PWR if you have the steam generator available the steam generator is the best the best system to remove to remove the heat so you don't use when it's possible especially when you have still high pressure in the primary circuit you cannot operate in several cases the heat removal system it starts working only with the pressure because it's a system designed for lower pressure it's not designed for the pressure of the coolant system so you tend to remove the heat to depressurize using the steam generator and some design have the possibility have an auxiliary feed water system for this situation and this is called auxiliary so it's not a safety system but in some in some design is auxiliary feed water system so the system that feeds the steam generator with water to produce a steam is a safety system so it's a function of a safety system so it depends on the strategy and in the PWR in some situation you have the possibility to discharge the steam directly to the atmosphere because the steam the steam generator is clean it's not contaminated with emergency feed water system and discharge into the atmosphere and then when you have reached the condition to which the RHR system can start working you operate this and then this can operate for longer term and you transfer the heat to the heat sink so emergency cooling of the reactor core when we are in the accident reactor core shall be provided to restore and maintain the cooling of the fuel under accident conditions at the nuclear power plant even if the integrity of the pressure boundary of the primary coolant system is not maintained so we are in case of loss of coolant accidents so we are explicitly addressing the emergency core cooling systems and this means provided for cooling of the reactor core shall be such as to ensure that cooling of the reactor core compensates for possible changes in the fuel and internal geometry of the reactor core cooling of the reactor core will be ensured for for a long period of time of time of time of time of time of time of time for a long time design features such as leak detection systems and we are moving another topic now appropriate interconnection and capability for isolation and suitable redundancy and diversity shall be provided to fulfill the requirement above so these are specific aspect of the design of this system it is already mentioned several times here as an example because you know in our safety standard there are no numbers in the safety guide some cases we have numbers but we are not allowed really to use number in our safety standards because this number can be different and are different from member state to member state and so we should we should write things that are applicable practically to all member states so this makes our life difficult so if we have to go to some really acceptance criteria and give numbers we have to refer to other standard or practices of other countries but this I mention here because this is adopted practically by everybody comes from originally from NRC but now is everybody I think is using the same criteria so the peak cloud temperature to be reaching that we are allowed to reach in a design basis accident condition maximum is 1200 degrees C the maximum clouding oxidation the calculated total oxidation of the cloud it shall nowhere exceed 0.17 times the total clouding thickness before the oxidation then we have criteria for the maximum hydrogen generation of course hydrogen generation is related to the oxidation because is the main mechanism to create hydrogen so something on the coolant geometry the coolant that the core should be still coolable and this is Tony mentioned this morning long term cooling after any calculated successful initial operation of the ECCS the calculated core temperature shall be maintained at an acceptably low value indicate shall be removed for extended period of time required by the long lived radioactivity remaining in the core so this is an example of practical value then here I have some drawing that a little more modern reactor rather than what I selected at the beginning so here you see the typical reactor coolant system of two loop reactors you see the system to depressurize the reactor two loops and here there are the safety injection system with accumulators and pumps different value of pressure to face different accident or different moment of the action here we have a system to remove the heat spray inside the atmosphere the containment and recirculate the water so is the containment heat removal system and here we have what we have mentioned before the emergency feed water system so this second possibility to feed this is the main feed water that comes from the condenser this is the auxiliary and the steam can be discharged to the atmosphere through this system of valves then there is something else yes so these are this is the cooling chain with the loops and they go to the to the ultimate automated sink and this is the the core, the volume and the chemical control system this is another example but you can find plenty of these examples and then you see if you the more you go into detail and the complexity is increasing and in reality the systems are very complex this is for typical passive system like AP1000 and that is a final consideration how to see what how the design the design is affected by the safety analysis of course you start from a design maybe it can be a preliminary a preliminary design it doesn't work anymore you model your plant for your computer code this is a typical modeling of a PWR in a computer code like a rail app this code for the fluid analysis then you determine the characteristic, the value of the parameter that you have for different accident and then you can correlate or implement other systems in your design so the final design is a result of many cycles of analysis and fixing and improving the systems so I think that's it so this is just the summary what we discussed the design of reactor coolant system associated system very considerably among world's population nuclear power plant because we are entering in the specific technological solution so we have to deal one by one assessment are relevant part of the safety analysis report this because all the we have a dedicated chapter for the for the coolant system normally chapter 5 the coolant system alone then we have in another chapter we have the emergency core cooling systems in the US standard 1.7 therefore reactor 1.7 is in chapter 6 it's called engineering features so part is dedicated to the core cooling system and part is dedicated to the containment and so on, so these are spread are spread in different chapters but all these systems are addressed in the safety analysis report the design of the reactor coolant system associated system of modern reactors is affected by the latest design requirement independence of level difference in depth in particular and the introduction of the design extension condition so the including the severe accident in the design and multiple the result of multiple failures even in safety system more than one failure in the safety system and by the extension of the life of the plant entity what we said before because now the trend is to operate the plant for longer time this is finished was a long discussion long concept that nobody is sleeping that's a very good sign maybe the food was not so heavy today why one is better than two I don't know this is what you don't understand how can one system be more reliable than two systems in two different situations you can decide case by case but I think the good solution is to have independent system for independent function of course you have to consider all the situation you can have probabilistic study if you conclude that your solution is better you can implement that but I have my doubts I am convinced that you can demonstrate everything with PSA you just fix what you want to achieve, you make a PSA you tune some values they can demonstrate this exaggerating I hope there are no PSA experts because but of course at the end is the analysis that dictates what is the best solution if you lose a system in normal operation and you need the same system in an accident caused by the failure of that system, what do you do then so this is the point because the system in normal operation can fail and cause an accident and you want to cope with the accident with the system that already failed so this well then ok if you have four trains like four systems yes it depends of course with probabilistic safety analysis you can make this decision of course you can increase the number of trains you can feed these trains with independent source of energy for example so there are different solutions there is no single solution so also the way you can realize the independence can be different for example using different power sources is already a way to achieve this independence you are going into the direction of increasing independence I think one is one is a I don't know which flight was this the first one the first one let's see if I can something can be also wrong if you don't it's not science which slide you see the first is B was in the pressurization thing B was in the this one back this one it's not too good and we will read it about favor I mentioned before the over pressure system is a system that is like the safety valve or can be the relief valve to control the pressure this fast the pressurization system is dedicated system to severe accident that we mentioned what I intended to say there is just the system to depressurize to avoid high pressure core melt situation ok and also in this case some design they are using they tend to use the same valves with the two different functions but also this is also when you are using the valves for different for different plant states maybe the valve they have to operate with different condition of the fluids so you have to qualify this valve for the most severe situation so it's not immediate that if a valve is good as a safety valve can be also good for fast the pressurization so but the you see the over pressure protection system is an automatic system always if you see the threshold opens the valve the depressurization system for this purpose is manually operated the operator that decide when he has all the indication that there is nothing else to do so cannot be exactly the same system or can have some components in common but also this is a bad practice yeah yeah yes from function also this is true yes of course this depressurization system is a feature is a feature for deck for design extension condition with core melt not well if you assume that you are a core you have a molten core and you are going to reach a molten situation of molten core you are in severe accident we call severe accident those involved in the core but when you decide to depressurize maybe the core is not completely molten or part because the core is not instantaneously the melting of the core takes some time you do everything you do but when you realize that you cannot remove the heat anymore and you are going to melt the core that is what you do Marko, let me explain it in a very simple way why do we need the depressurization front depressurization system severe accident condition when you have the core melt at a time at a time moment when the reactor vessel fails at that time the diamond containment in the variable core so it will be an explosion we have lower pressure in the system when the reactor vessel fails but it will simply separate from the one part from the other without any explosion so when we can already avoid what we need and then that is why we need the depressurization system for the accident condition the overpressure protection system is a different system, it is simply prevent the overpressure of the failure the main pressure boundary which can cause some damage of the reactor so these are two different systems one is for the diamond accident and the normal operation the other is for so the most important thing here is that at the moment the vessel damage due to over heating of the core mine should have low pressure in the system otherwise it will be an explosion and we don't know the consequences the consequences are unpredictable therefore we want to avoid it so this is what the depressurization system is doing about the normal safety valve normal safety valve normal safety valve can be a different function so normal safety valve is to protect against the overpressure so usually they are automatic the depressurization should be in fashion so you have to therefore you have to have a system which can be operated manually and it is easier to separate these two functions with two different systems than to incorporate these two functions that will complicate the depression complicate the design and also you will have to avoid the unintentional depressurization during normal pollution so if you incorporate these two functions into one function hopefully by one system you have to be careful with the design because it may not have some unintentional depressurization during separation during normal pollution which will be an aspect of initiating the design which will be something like design design which will be an aspect of initiating the design which will be something like design design so it depends on the design because if you don't put in the pressurizer you have to put new penetration either in the vessel or in the piping so this also so normally all the solution I have seen they are in the pressurizer in principle can be also somewhere else there are designs yeah, bad designs let's say and even I know that in design we also try to some of the regulations require to separate the function we separate by the regulation of the function and we require as as function and how do we operate the pressurization system with the same DC power the same compressor how do we operate this or you have a dedicated system in Hungary and then ok yeah ok because one thing is separate the valves you have two different set of valves but the valves are only part of the system so the valve you have manually operate you need the system to operate can be compressed air, can be electric power DC power so if you want a complete separation you have to separate also the source of power the question did not appear before for the order of plants because we concentrated on the design this accident and we did not have any equations when the core has already the core value should be same as now in the new design there is a requirement to be with the core value as well as within the design in order to avoid large relief to save the container to save the container against over pressure or fast pressure and that's why these new functions come into the picture in the new design to cope with such phenomena which we didn't concentrate before ok thanks other things but I think we are already 10 minutes later probably better to then we can have some question at the end of the session in the discussion