 Мы опубликовали босс на STP-сетях, на IA-сетях, и это и среклассно реклассно. Это будет, как они говорят, на YouTube-сетях. Мы будем организовать все материалы в один, в том числе видео, все обучения, и все ваши посты, и все листы патисментов, и вынер и вынер из постер-сессии тоже. Если вы хотите победить, пожалуйста, подпишитесь, потому что если нет, то нет. Окей, давайте продолжим сейчас. Как я сказал, основные симуляции, как балансы, температуры, флорады и другие основные симуляции, они очень основные, и могут быть сделаны довольно просто, даже с помощью номерикалов и севдекодов или субчаналанализов, но также с руками, потому что первые реакторы были калкулированы нередко с руками, или с логарифмикой рулёрой, или с предыдущим калкультером, даже с севдоками. Однако, когда оно идет к транзентам, оно становится гораздо сложнее. В транзентах есть различные типы транзентов, например, реактор стартап, который очень слегка, может быть считан как сервисы из Several Steady States, но есть эксиденциальный транзент, который является результатом эксиденции, и это включает основные эксиденции. ДБС, мы говорим, когда реактор стартап нормальный, или он может быть дровом или релизом на сингл-контаро, что result in the power excursion, и также loss of one of all primary pubs, but it is a protected accident, which is with normal shutdown. However, this is what is included in the design and must be simulated in the design stage, but in addition also for simulation in the design stage, we have so-called design extension condition accidents, which includes, okay, there are different terminology, I would not go to the details, but generally this is this accident, that are severe accidents that may result in core melting, or core damage, or core disruptive accidents, what we say. And that what we want to prevent, or in generation 4 reactors we want this, or innovative nuclear energy system, we want to practically eliminate this possibility and with core melt and practical elimination means to create the design that physically eliminates the possibility of such event, or when it is physically impossible to make the design measures, which will lower probability of this accident to the negligible values. Not zero, but something like very low probability compared with other potential effects, like meteor or, I don't know, which is, and then we call it practically eliminated. And it's one of the goals of generation 4 international forum, generation 4 reactors. Never thought, I mean, no, not never thought. For this reason we also should simulate, even those possibilities are practically eliminated, we should simulate also those transit, which may or which result in core disruptive accidents, just to see what can we do to prevent it, how to mitigate it, and what potential consequences it could be. Of this is the most important and most dangerous, it's so-called ULOV, unprotected loss of flow, and it's most serious accident, and it means you have, you lose all your pumps, and the counter roads and shutdown roads don't work, so shutdown system does not operate. That's what happens with the reactor. Another severe accident of this type, I mean design extension conditions, is unprotected trip of power, that could be related to the sudden, for whatever reason, drop or release of the, or full bank of the counter roads, so like all counter roads are out of the reactor, what's happened next. And another type of the accident, I mean, I'm counting only main one, it's a core flow blockage. For whatever reason, for example, we have some, let's say, we did not control oxygen carefully, and we have some pieces of oxides, which are collected somewhere, and then block the inlet of the fuel assembly, and that will reduce the flow rate, whatever, significantly, or even totally block the sub-assembly, then, if you totally block the sub-assembly, what's happened next? It will melt, of course, inside, the cladding will fail, and the fuel probably will melt, but will it distribute it more to the other sub-assemblies, or it will remain local accident that can be mitigated as a result. So, and the loss of heat sink also, let's say, for example, for whatever, like, decay heat removal system don't work, it's also kind of accident, and it's also transient, but, let's say, they're all in the different time scale, when your love is very quick, it's several seconds, for example, for SFRs, and for, like, most of heat sink can, and several hours, days, and so on. So, to solve this problem, we will need, it will require coupling calculations from hydraulics, nitronics, mass transfer, a kitchen of states, and probably some structural material calculations, which we need how materials can survive this accident or not. This is very complicated, multi-physics problems, and for this, we have several computer codes developed, and I just want to show you several examples. Of course, we have no time to discuss all the details, but I can show you several examples. I took, let's take as an example of the Astrid, former Astrid reactor, which was presented by Christian two years ago, here in Trieste, and this is very advanced, definitely one of the leading generation for concept, even the Astrid is not officially part of the sixth generation for concept, I believe, ESFRs of this. Anyway, this is innovative, sodium cooled fast reactor, which satisfies all criteria for the generation for reactors given by GIF. So, you have, it's a pull type reactor, here, and we have the, so reactor is here, vessel, relatively big vessel, inside we have a quarry, which is, I will show you later, then you have this vessel, and here we have a pump, which pumps coolant from the so-called hot plenum on the upper part to the cold plenum, and then we are, the flow goes inside the reactor, which is released as usual in the so-called cold plenum, because coolant already hold, then it is sorted in the heat exchanger, and from heat exchanger it is released already cold to the cold plenum. So, it's not directly connected, like outlet and inlet of the quarry with heat exchanger and pump, but you have heat exchanger and pump actually in different positions, slightly different positions. So, it's also, it's a little inertia to this system, so it's not immediately react on, according to the temperature, does not change very quickly, it remains there, and also, okay, then we have this intermediate loop, steam generator, or maybe other system, like conversion system to the electricity, and so on. So, here, we have innovative quarry designed for enhanced safety features. As for the standard, problem as we're discussing already, it's a sodium void reactivity effect. And there are several design measures, assumptions, and many numerical simulations how to reduce this sodium void reactivity or eliminate, or practically eliminate the possibility of the severe accident in such a quarry. So, this part is shows the example of the simulation of the Astrid. It's the same picture from the previous slide, as it's simulated with SAS-4A and Simmer-3 codes. Simmer-3 is two-dimensional safety analysis code, multi-physics. And in this case, we use two-dimensional version, because three-dimensional would be too much for this system. You see, the quarry is again here, we have somehow simulated pump, heat exchanger, this upper planum, upper structures, and the die grid, and the inlet is here. This shows the temperature distribution and the nominal conditions in this quarry. You see, I will show you maybe more details for the quarry. So, it's more or less simulated and the nominal conditions. Now, what happens if you have a UFO accident? Before we go there, let's explain a little bit about the enhancement of the safety, which was done for this. This is one of the versions of the Astrid. It's not, of course, final. The feature was published like five years ago. We can see. So, the features of this quarry, as you might see, we have a quarry with different sizes. In this case, it's homogeneous, but it has different sizes of fissile fuel assemblies here. And also it has extended sodium planum, especially here it's extended. And the sodium planum is extension of the sodium planum, the purpose of this sodium planum. So, it's like more or less empty subassemblies on the top of the outlet. It's due to the leakage, reduced. So, if you don't have sodium here, it will actually have negative reactivity effect. Because in the central part of the quarry you have actually always, you always have a positive reactivity effect, which you want to avoid. But in the upper part of the quarry we don't have fuel, but we have sodium. And if it's sodium voided, we have reactivity effect. So, finally, total reactivity effect is positive. Even this plus one, and this should be like about 500 pcm positive. But we don't expect that the sodium boils immediately, removes from every part. So, what we expect, and what's the purpose of the upper sodium planum, that since we have this also negative reactivity, sodium will boil in the upper part because temperature in the upper part is higher than, it's maximum temperature of the coolant. And pressure is lower than. Plus, also, it's a little bit different altitude, I would say, height. So, pressure is lower. And also, pressure is lower during flow, coolant travels through the quarry. It loses the pressure. So, minimal pressure in the upper part. And since it's minimal pressure there, so boiling should, in principle, start in the upper part. And if it starts in the upper part, which has negative reactivity, then we expect negative reactivity, will lead to the reduction of power and shutdown of the reactor. So, assimilation is done for this. This plot shows flow rates through the quarry versus time. So, as you see, it's very quickly, not very quickly, but relatively quickly, actually in few seconds, pump stops. But then, this flow rate becomes like about 10%, after some time, 10% of the nominal and remains of the nuclear circulation only. It's 10% of the nominal flow rate, provides nuclear, sorry, natural circulation effect. Then, when, in this case, as you have very low flow rate, power is nearly nominal, a little bit less than nominal, you have, obviously, overheating and sodium starts boiling in the upper part. Once it starts boiling in the upper part, you have decrease in reactivity here, and you have decrease in reactivity and decrease of the reactor power as well. Then, sometimes boiling stops, because it becomes colder. Boiling stops, in this case, sodium comes back again and you have again 740 plus P7 of the reactivity. For some calculations, it should be like a periodical process like this shown on this paper. So, it boils and stops, boils and stops with the power a little bit even close to the nominal, no, sorry, it's reactivity, it's reactivity flow. But there are several other cases and they are very sensitive, it shows sensitivity and I will show you probably on this page now. So, here shows the reactor power. This here, it's logarithmic scale, so actually power becomes like maybe 30% first thanks to the Doppler and something, after the trend it first becomes a little bit lower and then this process starts again. And let me show you. It should remove like... Okay, thank you. So, this slide shows the simulation. So, we see time goes after the accident and then you see in this area the boiling starts, we have empty voided area. Then after some time it stops because it becomes the reactor power lower and becomes colder. But later it's again. So, sometimes you can see interesting effect like it's stable, boiling stops starts like periodical process. However, it's very sensitive. If you change parameters, in this case it was parameters of coupling between SAS 4A code and CIMER code, you can have either immediately secondary power excursion which destroys excursion very quickly showing power, you see, in red which could be up to 500 of the nominal power peaks. Or you can have like stabilization of the process simply without any boiling on this shown in the blue line. I'm showing this because it shows how complicated could be this effect. And also especially if you do it for the new reactors and you try to find out whether this still numerical simulation shows that several possibilities possible because you see this sensitivity analysis shows that we don't know exactly. We cannot say what would happen in the reality and we should consider all the variants and unfortunately in this case we could, maybe it's not very realistic but we should be conservative in this case that it could happen that still this will not help us and other things. However, it's these results were like it's not how to say the final answer of course but because we don't know the code should be, you cannot validate such systems complicated on the real situations because we don't have experimental data and I hope we will never have such kind of experimental data for the fast reactors or whatever reactors as well but this is very complicated problem that cannot be solved and we should work and then on the design stage we should demonstrate that our reactor can do this and this and if in case of accident we could this so it was actually wrong timing so it should be for the coffee break. Okay and to help our member states in validating and verifying their numerical simulation codes and models on the regular basis so-called CRPs or coordinated research projects or benchmarks on different aspects of this kind of system and most of them are simulations of the reactors, transit and so on. This slide shows recently finished in 2016 benchmarks that was CRPs that was conducted for years its benchmark analysis of the EBR2 shutdown heat removal test so that time on the EBR2 it was a relatively old reactor it's very interesting and very complicated system it's mix of pool and the loop type actually it's pool type reactor with coolant circulating in loops inside I would say and from the lower part it's actually the pool type connected through the Z pipe so it's actually loop type reactor so the purpose of and they conducted this test in INL to prove that reactor could survive CV accident it was like maybe 40 years ago or more Finally those experimental data became available and INL decided to open to provide the data for the community and for the IEA and the purpose of this therapy was like it happens like okay, first we define the configuration define the experiment and whatever happens whatever had thermocouples what was measured, what was not but also the initial conditions boundary conditions as much as possible and then participants start blind calculation simulation so they model the core whatever model they have with whatever simulation code they have then after let's say about one year of simulation they gather together again and compare the results compare the results for between each other and versus experimental data then at this stage the experimental data the results of the experiments are disclosed and you can compare and here this I put just on one slide several slots so you see this specific test shut down test number 17 which was connected to the it was protected loss of normal and emergency pumping but was protected and you see this like many about 20 or maybe I believe it's 15 simulations of this code and the results are compared with each other and with experimental data as well so then basically after that first comparison we would refine their models maybe they would choose another code or they introduce another model or update their models what they are using in the simulation of such accidents and then compare the results finally again and we have like document which describes the experiment what was done and actually this document provides a good very good base if somebody want to validate a new code a new model they would take this document repeat the simulations and compare and that would be a very good case of the validation that how we try to improve and share the knowledge on this technique on this complicated sophisticated computer code methodology for our member states that is another slide on the bar shutdown it's another it was one protected loss of flow and unprotected loss of flow but however it was screen was disabled station block out but it was an additional pump actually pump still was working so it was not classical it was complete loss of coolant but it was some calculation so and this slide again shows experimental data calculator for different temperatures and then again one example of comparison of the participant data of UC we have many calculations very wide range some of the source of the discrepancy could be simply error in the input data because it's very complicated system as I say you can you should model very carefully geometry and other many many other things and any mistake of course you should different codes use different models even within the same code you can use different models different machine whatever there are many many things that can affect the results and the task this was to compare not to say that this was not to let's say assign the best available performing code in this case of course but it was just compare and give the participants possibility to improve to refine their models and also to give the further let's say to share this all experience with other possible participants but for the people who are developing such codes and need these experimental results for validation of their codes and that was successful CRP which was published and I believe it was very successful CRP everybody were happy especially under the INL Argon National Laboratory was really leading this CRP organizing and it was very good example of organization and join work on this benchmark problem now we started two new CRPs this year and one of them is another which is called which mark analysis of FFTF loss of flow without scrum test it's how they call loss but usual name is you love but since that original test was loss of flow without scrum so we did it keep this name for this CRP but it is simple FFTF is fast flux test facility in the PNL now it is not in decommissioning but still innovative but can be restart operation on this time but it was stop like 20 years ago about and like 30 years ago they had some experiments so this reactor was a prototype of the SFR which is 400 megawatt thermal sodium cooled fast reactor of the pool type because low power is lower and it works on the mox fuel mixed uranium with dioxide and it's loop type with axial and radial reflectors and the total volume of the core about 1 cubic meter is small and with 90 only centimeter height and diameter is like 120 actually fissile core is very small for the fast reactors and one of the features because we need high neutral flux and we need high power options of course high power density it's several times higher than let's say water reactors and one order of magnitude higher than let's say high temperature graphite reactors and so on and to demonstrate the safety features of this reactor, FFTF they did several experimentals on the loss of flow without scrum test and at that time they already implemented okay passive safety system which is gas expansion model interested system like say if you have lower flow you insert more gas and I will not go into detail but the purpose of this particular test was to demonstrate that this passive shutdown system works well okay and so the test was it's one of the many tests that are on this particular test the reactor was working on the nominal power or 50% of the nominal power and main coolant pumps were turned off and normal control scrum response was disabled so then more or less you see the results of the experiment okay this is not so and temperatures were reached some somehow the maximum values but it was still within the margins of the safety limits and thanks to this gem gas expansion model passive shutdown system reactor was safely shut down without exceeding safety margin however let's say notice it's at 50% power and I believe if they start from 100% it would be not very successful in this case of course they had some let's say if temperature would reach some you know unsafe values or limits they would shut down the reactor anyway but in this case reactor was not shut down I mean by control it was shut down automatically but thanks to this gas expansion model passive system so more or less you see here shows the coolant temperature it was rising and then going down and comparison with Rhapsody which was by the way they compared with something and the result is that coolant temperature never reached the boiling limit and it was I mean good demonstration so now it's and that we don't have boiling it's very good because once you have boiling the solving of the problem becomes much more complicated and very oops, sorry and it's uncertain and the result becomes very uncertain because it's very complicated physical phenomenon in this case and never the less even in this case it's very complicated task how to simulate this system properly accurately using only this input data and we had already first meeting that we will have but already we have specification and in October we will have a first meeting where the participants which is about 25 organizations from 15 or 17 countries are participating now when participant will discuss decide what they actually need how to model to find some general common solution and then go for their blind simulation for the first year and just in addition I would like to show you another CRP which we started it's not ULOV and not transit analysis but it's Neutronic Beachmarks of CFR startup test it's pure Neutronics where people will use Monte Carlo and deterministic codes it's standard task relatively easy I say relatively because it's it's not easy at all but since many countries many organizations and many participants have great experience in simulating of such kind of let's say only Neutronics you don't have to simulate coupled with some hydraulics or whatever so this I expect if we could receive a good results and if you need experience in Neutronic simulation and many people let's say and I believe your organizations you are presenting are also several participating in this CRP it's another example of CRP we have and what importance already Chirayu on the first day or he explained that we are to you know that the idea of this CRP is to collect data experimental data to have a benchmark and then to provide for the future users this set of data so you can repeat this experiment and simulate this experiment and compare your results future models with this data however if you have only book it's very very complicated specification it's not always easy that's why we are in support of this let's say participants now and for future they develop so called digital nuclear reactor it's the storage of the data on the core on the reactor which helps the users to calculate and we have this already for the CFR startup benchmark and we are planning to have it also for the FFTF reactor core at least I just want to note one more thing which I've forgotten in my first presentation and maybe useful for you within the NAPRO CRP sodium properties we will issue not only text books or hand books on sodium properties including physical and chemical sodium properties and friction factors and heat transfer coefficients for the sodium coolants but also online sodium properties calculator when you can there is a working version maybe we can also check how it works it allows you to use all relations included in those handbooks will be also included in the system so you can easily see the results obtain the result see the comparison and also you can extract the computer code I mean the code exactly which can be used in your models for the modeling in your computer this is Java I believe and it should be also easily converted to C programs as well ok thank you very much for your attention and thank you very much