 So, the next topic is the analysis or assessment of safety over the lifetime. This is an interesting topic because this can create most of the headaches for the operators because they have to spend money on safety assessment once the plant has already been licensed and they start the operation according to the license and everything and then comes the regulatory body saying that, hey guys, do you follow your safety assessment and do you update your safety assessment with the changes and everything? So it is then the plant has to organize some projects which fulfills the tasks of the requirement in relation with the requirement. So, unfortunately, at the beginning of our nuclear power plant operation back in 1980, 81, 82 and so on, such requirement didn't exist and then the first so-called periodic safety review should have been submitted to the regulatory body 10 or 12 years after the start-up of the plant and at that time we in the country did not have regulatory guidance on the content and how to perform such periodic safety review, so-called periodic safety review because at that time already it was called so and therefore the plant management or the plant technical stuff, technical management and the regulatory technical management sit together to agree on some kind of content. After that, of course, it was after the first year period which included also so-called re-evaluation of the safety, which I have already talked about, that was a complete re-evaluation of all safety, producing all the safety analysis including the probabilistic safety analysis and the deterministic safety analysis together with the severe accident analysis. And once it was done, after that, after 12 years of operation, we can say that the flow of the normal life went into the right track. So after that, the periodic safety review was required with the periodicity, predefined periodicity, so not on a not whole basis. So the assessment of the safety over the lifetime of the plant and the periodic safety reviews, this is the presentation title. Now I will have some introduction design lifetime considerations because this is also a topic here. Then the modifications and design changes, how to take them into account and what to do with them. And finally, the periodic safety review reports in relation with the periodic safety reviews. So well, we know that the nuclear power plant safety is changing by the time. It will not be the same as at the beginning of the operational cycle or not at the beginning but sometimes after the burn in period when already the number of transient events or number of reactor scrims already stabilized. Those who already started the operation of a nuclear power plant knows that at the beginning of the life cycle there are lots of reactor scrims and then the number of reactor scrims is lowering by the time and we were the same. So for the first unit we had, I don't know, maybe five reactor scrims in the first year of operation. Today we have one in ten years of operation. So this is the real effect of the maturing of the plant. However, there are other factors which are affecting the safety. Those are the modifications and design changes. Not in every plant there are such changes. In our plant we had lots of changes, lots of safety, so-called safety upgrading measures implemented. It was something like 60 or 70 safety upgrading measures were implemented since the beginning of the lifetime of the plant and each of them affected the safety very significantly. On the other hand the aging of the plant components is a factor which may increase the risk of the plant which may somewhat degrade the safety because there are aging components. By the time of course if you are using some components the wear out of the components causes the aging. So that also creates a problem with increasing number of equipment failures or increased frequency of the failures. With the increased frequency of the failures the unavailability of the component will be larger. Therefore the time when the component is unavailable is increasing by the increasing number of failures. And there are organizational and personal changes, definitely, yes, yes, yes, yes, that's right. Okay, this is a good question. The answer to this is that what we are estimating the aging. We have to follow the aging, whether the aging goes like it was designed or it is faster or lower. So in many cases unfortunately the aging may be faster than what was designed. And it can create problems with the lifetime of the equipment. So the actual design lifetime of the equipment is not exactly what it was designed for. Let me tell you one story with the reactor vessel of the VVR-440 reactor. So now today it is a well-known problem that the reactor vessel was very close to the core. And the fast neutron flux at the edge of the core was rather high and the effect on the reactor vessel wall was causing faster embrittlement of the reactor vessel than it was designed. So at the beginning of the operation this effect was not fully understood or was not fully known. For instance in Finland they already had to do some measures to recover the core, I mean to recover the reactor vessel to be able to operate through the lifetime of the reactor vessel. So there are some factors which accelerate the aging of the components, therefore it is very important to follow the age of especially those equipment which are important for safety. So now organizational and personal changes, well management changes because of aging of the people, 30, 40, 60 years of operation is long enough to have the personal change as well. Only new management will start their activity with changing the organization to fit into their, I mean to fit their taste. This was also in our case, there were several reorganizations of the plant personnel, there were several changes in terms of organizational units and of course there was a trial with the increasing cost of the, increasing labor costs, increasing salaries, those costs become significant contributors to the cost of the electric power production. Therefore there was an intention to try to minimize the number of staff at the plant and this is very much like in the other countries as well. So at the beginning we had like 5,000 people in the staff for four units, now we have two and a half or something like that, which is half of that. So these changes definitely will affect the safety. So because we have less people, we have different people and with the change of the people there are different knowledge, different training and so on and so on and there was a time even when we did not have the training, I mean education in the nuclear field because it was not needed. The market was saturated with nuclear experts, there is no need for newcomers. After a time it comes to a problem because people are also aging so they go retiring and then comes a problem, hey, there is a new young generation and we had to establish, we established in Paxch a school specialized on educating the new generation and today we get rid of that problem. So this is already a question of the knowledge management, which is a very popular terminology these days, knowledge management. The olds are going out with their experience, with their knowledge and nothing remains, just the documents. So this is a difficult issue and I guess most of the countries experience these difficulties if there was no education in the specialty of the nuclear field. So organizational personal changes are affecting the safety. Many of these changes are subject to licensing, which means some of these changes are already licensed by the regulator or are subject to licensing by the regulator. So you cannot do anything, I mean everything with your organization, this should fit into the operational needs and this is very important to assure the necessary number of and necessary qualification of the staff as we talked about it before. Then how funny the safety is changing with the development of safety analysis techniques, how it comes. Well, we analyzed something at the beginning and with the increasing knowledge and we are also developing with the knowledge, we have new techniques, we are discovering new phenomena which we haven't thought about. What we thought was safe before, it is not safe anymore today because something become clear which was not known at that time. So the safety analysis techniques with the development can show some new problem areas. So we are analyzing on a totally different way the safety problems as let's say 30 years ago. So it will happen with new plans as well, especially if we are trying to cover a 60-year period, just imagine 60 years. What was 60 years ago? Could you imagine the development, what has happened during these 60 years and then comes the question whether we are foreseeing that development for the next 60 years. Therefore it is important to follow the developments and do the reassessment of the safety with the new techniques. Changes of regulatory requirements, do they change? Of course they do, with the development of the safety analysis techniques with the development of the overall knowledge. Of course there are some political changes and how to say, public relations changes. The public, how funny it is, earlier we had less safe plans but the public acceptance was better than today when we have safer plans. Come on, this is a contradiction but this is true. So what can we do to satisfy the public? We need to increase or not increase but strengthen the regulatory requirements and the regulatory requirements are getting more and more strict by the time. Just a good example. We have different regulatory requirements to the old plans as to the new ones. By a factor of, by one order of magnitude we are lowering the frequency of the allowed, the maximum allowed core damages, the maximum allowed frequency of the core damages. We are lowering compared to the old nuclear power plants, one order of magnitude. So for the old reactors we are happy if the overall core damage frequency is somewhat 10 to the minus 4 per year or below 10 to the minus 4 per year. It would be unacceptable for new plants. We are expecting less than 10 to the minus 5 per year core damage frequency and the frequency of large releases we are, we also lowered down below 10 to the minus 6 per year, which is almost once in a million year. Come on, that's something which is, I mean we cannot imagine, once in a million years. So sooner or later we'll go down to once in a billion years and then it will be once in the lifetime of the earth. So that's which is already in the region of unbelievable. So this is how the progress is going. Today we have some set of regulatory requirements more or less harmonized on the international level. There are some discrepancies or there are some differences country by country, but more or less except the U.S. and France and a couple of big countries with large nuclear industry. The requirements are driven by the IAEA standards. And this is good because we can have sort of a common platform for the discussions on the international forums on this. Now due to new planned design developments and public expectations, I was talking about the public expectations, but it is interrelated with the strengthening of the regulatory requirements. And new planned design developments, just we had something at the beginning in terms of so-called seismic design of the plant. Seismic design considered an earthquake with the frequency of 10 to the minus 3 per year, 10 to the minus 3 per year, once in a 1,000 year. And then the general international requirements increased, I mean decreased this number to 10 to the minus 4 per year for the old plants. And it created a headache to the plant people how to strengthen the plant structures and components to survive 10 times, I mean earthquake which is more intense than the one which they were designed for at the beginning. And if you go to Pax and you see the plant, you will see some funny structures which were not there at the beginning around the plant buildings which have a role to strengthen the seismic resistance. And also inside the technology there are strengthened supports and some new structures built in in order to increase the seismic resistance. So today for the new plants it is not necessary because they are already satisfying the present requirements. I'm not sure about the future requirements because who knows where the future requirements will go. So it is necessary to assess the safety and demonstrate the acceptance of the safety time by time during the lifetime of the facility. That's clear. We are increasing just if you think regulatory requirements are changing, then you have to demonstrate the compliance with the new regulatory requirements and full stop. It may require a full re-evaluation of the safety as it was in our case. So that's something which is to be considered already at the beginning of the operation during the design. And as we have the design lifetime considerations, what we discussed just a couple of minutes ago, the design, during the design, the aging of the components has to be taken into account. Now the design lifetime of the plant is dependent on the non-replaceable plant component with the shortest design lifetime. I will repeat this in the next presentation as well because this is very important to understand what does it mean. So non-replaceable plant component with the shortest design lifetime. If it is the reactor vessel, because we are not replacing the reactor vessel, then this will limit the lifetime of the plant. If it is the steam generator because it is not replaceable, in some plants it is replaceable. But in the VVRs, for instance, these big horizontal steam generators are not replaceable. There is no way to replace them. So no technology to replace them. Therefore if I'm doing something with the steam generator which will end the lifetime of the steam generator, then full stop. This is the lifetime, end of the lifetime of the plant, of my plant because I'm not able to replace it. So this is how to understand the lifetime of the plant. Now, of course, for those components which are non-replaceable, I have to establish at the plant some control program which will define at any time the residual lifetime of that component to know where I am in time. So sometimes the material investigations, the investigation of the samples taken from the reactor to understand the progress of the embrittlement of the reactor vessel material may show me that, okay, so if the progress is such fast, then I have, I don't know, maximum 10 years of residual lifetime. After that I will have to stop the plant and so on and so on. And this is the same with the body of some pumps or the steam generator vessel and the steam generator structure. So that's why we are making, we are doing time by time some material investigation, okay. These are the so-called in-service inspection practice, okay. Now the design progress of aging effects of the non-replaceable components predetermines the lifetime of the plant, definitely. If the lifetime is 30 years, of course with the margin, there must be a safety margin which allows me to run the plant for 30 years. This is of course something important. But then I am investigating, I am doing such, or I am designing such core, let's say such core which has so-called low leakage core design. Low leakage core design means lowering the fast neutron flux at the edge, which is, if I am lowering, then the fluence of the fast, fast neutron fluence causing the embrittlement will be smaller and it may help me to increase the lifetime of the reactor vessel, okay. If it is the limiting factor. So and this is what we did, therefore the embrittlement didn't progress so fast, therefore we could increase or extend the service life of the plant. We licensed 20 years extra operational life for our plant. But we had to do these investigations and we had to justify that this allows that. Also allowed by the design maximum number of the cycle, cycles of specific events may be a lifetime limiting factor. I mention it especially, we try to not to concentrate on the allowed number of cycles, load cycles, but there are load cycles. Let's say if we are designing a plant to be able to follow the load, the load following characteristics, it will require to determine how many times I can change 1%, how many times I can change, I don't know, 50% and 20% and so on and so on, so that the operator was able to make the maneuvers necessary for the load following. But as soon as it expires, you cannot do anything. I mean it will stop your load following capabilities. So you can operate either only in base load operation or you have to stop the plant. That's it. So now we will come to this. It is essential to follow the development of the limiting factors to determine or to demonstrate the safety during the residual lifetime. Now the prime responsibility for the safety rest with the person or organization responsible for the facilities and activities that give rise to radiation risk, which means the operating organization. So the designer did his best, they will specify the life limiting factors, they will specify many other things, and then the responsibility of the safety will remain with the operating organization. So this is the operating organization who has to do those investigations which are needed to demonstrate the plant safety. Then the operating organization shall establish a formal system for ensuring the continuing safety of the plant design through the lifetime of the nuclear power plant. This is in the SSR 2 slash 1, and probably this requirement 3 has already been mentioned, but it will be mentioned another time during this day. So what about the modifications and design changes? During the operation of the nuclear power plant, the person performs technological changes and or organizational changes. Sometimes we have to. There are requirements to do that. There are some new developments which shows the necessity of doing so, or simply we have a better solution, or we discovered some weaknesses in the design. We had, again, I can give you examples on any of these because we have a nice experience with those changes, those design changes we had to perform. Like the one which was mentioned by Alec yesterday, that when the diesel generator starts, you cannot immediately load the diesel generator with all the consumers. You have to do some step-by-step loading program. You have to follow some step-by-step loading program. And that's okay. By design it is said so that if the large, large brake rocker in the primary circuit occurs, then the diesel generator will start. If we are losing the power, diesel generator will start. And in 20 seconds the diesel generator will be already on a level where the step-by-step loading program can start. Now it appeared that this 20 seconds is too long because after large brake rocker accompanied with the loss of off-site power, the pressure in the primary circuit will drop down below 150 degrees. So the pressure will go down so that the primary coolant temperature will be below 150 degrees. Which means another step-by-step loading program would start just because there are distinguished systems for the low power or shutdown cases and full power cases. So the system thought that the primary parameters belong to a shutdown state, therefore a totally different set of systems would have started. So it created, of course, everybody was surprised at how it comes to design deficiency. Let's do something. First question was to understand whether this event will be accompanied with loss of power or not. Then if it is accompanied with loss of power, then we make a delay with the consideration of what kind of step-by-step loading program should start. So we had to modify the logic of the step-by-step loading program in order to avoid such situation. And there are other stuff which comes from the experience of investigating the results of the safety analysis cases. So these changes may affect the plant's safety, therefore the safety of these changes should be justified and demonstrated depending on the regulatory requirements. Any modification at the plant has to be categorized according to their safety relevance. And for the higher safety classification or categorizations, let's say categories, not classification, not classes, categorization. So if I classify or categorize a modification to be category one, then it will need a special licensing process. So of course the categorization will depend on the safety significance of the change I am doing. The regulatory body requires, very strictly requires the categorization and the licensing process according to the category after the end. The national regulation should specify the licensing requirements of any plant technical or organizational modification that has relevance to the safety. These requirements cover also the safety analysis of the modification. So safety analysis requirements are also specified for such modifications. Now comes the question that do I reanalyze all the safety analysis after each modification? No. They are cumulated and then comes a time when I have to submit a so-called periodic safety review where I'm also submitting the safety assessment which considers the cumulated changes. And here comes the periodic safety review discussion where these are routine reviews of usually routine reviews of nuclear power plant operation including reviews of modifications to hardware and procedures, significant events, operating experience, plant management and personal competence and special reviews following major events of safety significance are preliminary means of ensuring the safety. This is okay. This is what we understand. Now some member states have initiated systematic safety reassessments termed periodic safety review. Some member states, more and more member states. Let's say so because not everybody is having, not every country is requiring such periodic safety review. The periodic safety review is assessing the cumulative effects of the plant aging and plant modifications, operating experience, technical development and sighting aspects. So we are re-evaluating also the sighting aspects which we were talking about yesterday. Yes, it was yesterday. The periodic safety review, the PSR includes an assessment of plant design and operation against applicable current safety standards and operating practices and has the objective of ensuring a high level of safety through the plant's operating life. Now this is what we are doing. We are reassessing the safety according to our actual plant status, actual environmental sighting, design and so on and so on, comparing them or justifying them, justifying the compliance of them with the actual regulatory requirements. And with this, of course, it will cause increasing safety over the lifetime of the plant. But it will cause, of course, lots of investment into safety improvements and into the safety analysis to be performed. Well nuclear power plant is producing electric energy so there is a source of these resources therefore we should not be, how to say, we should not take care of the cost of such periodic safety reviews. Then again, okay, so the safety of nuclear power plants, the commissioning and operation, this is SSR 2 slash 2 says about the periodic safety review which is a systematic safety assessment of the plant in accordance with the regulatory requirements shall be performed by the operating organization through the plant operating lifetime. We do account taken of operating experience and significant new safety related information from relevant sources. And the other safety guide which is issued by the IAEA is the periodic safety review for nuclear power plants. So this is the SRG-25 and it contains the content requirements for the periodic safety review reports and it contains the content of the periodic safety review assessment itself. It says although the operating nuclear power plants are subject to routine and special safety reviewed, such reviews are generally not sufficiently comprehensive to meet this requirement. Which means this requirement which said systematic assessment should be performed, this special safety reviews are not sufficiently comprehensive. This is what it says therefore a periodic safety review should be done. And this is the suitable tool to perform the sufficiently comprehensive assessment of safety during the lifetime. And this SSG or this specific safety guide specifies the rationale of the periodic safety reviews with a 10-year period. They say that 10-year period is sufficient. Why not five? Why not two? Why not three? Okay, we are not able to just continuously doing periodic safety reviews. The periodic safety review process is a long and time-consuming process, this is one part. And 10 years considered to be enough long to see the changes, to see the changes of the safety with the accumulated effect of the modifications during that period. And it seems to be short enough not to miss significant safety degradations. So this is the rationale and basically this 10-year period, 10-12-year period is accepted by the member states and it is followed. Now the changes in international safety standards, operating practices or technology underlying scientific knowledge or analytical techniques is maybe significant enough to be considered. The potential for cumulative effects of plant modifications to adversely affect the safety or the accessibility and usability of the safety documentation. Identification of significant aging effects or trends will also be identifiable, accumulation of relevant operating experience, changes in how the plant is or will be operated, changes in the natural, industrial and demographic environment in the vicinity of the plant, changes in staffing levels or the experience of the staff, changes in the management structures and procedures of the plant operating organization. So these are all explaining why about 10 years. This can have different purposes. The systematic safety assessment carried out at regular intervals, support the decision-making process for license renewal, support of the decision-making process for long-term operation. Well, the regulation in the different countries is very much different in relation with the license renewal. There are some countries where the license has to be renewed every year. There are some countries where the license renewal is switched to the periodic safety review. And there are some countries where when the operating license is issued at the beginning of the lifetime of the plant, it is valid for the whole lifetime duration. So depending on in which country we are, the purpose of doing a periodic safety review can be different. The objective of the periodic safety review is to determine the adequacy and effectiveness of the arrangements and structures, systems and components that are in place to ensure the plant safety until the next periodic safety review or where appropriate until the end of the plant operation. That is, if the nuclear power plant sees operation before the next PSR is due. So we are trying to assess the safety for the next period, till the next periodic safety review. We have to justify the safety, justify that the aging mechanisms are not jeopardizing the safety till the next periodic safety review. And the regulatory body can be sure that, okay, we accept the plant operation for the next 10 years till the next periodic safety review. Of course, if the plant lifetime expires within this period, then the safety should be ensured till the end of the lifetime of the plant. The other objectives, the objective of the PSR is to determine the extent to which the plant conforms to current national and or international safety standards and operating practices. Just as you can see changes in the IAEA standards, they are reviewed time by time and some new safety standards are coming out. Especially the question are related to the new safety standards. But as you could see in the last 30 years, how the structure of the safety standards at the IAEA has changed. So we are now renaming the safety standards to one, something was NSR, something now we have SR, SSR and so on and so on. So also the structure is changing, definitely the content is changing. Also with these changes, standard changes, the international acceptance is changing for the operating plants. Therefore we have to determine the extent to which the plant conforms the current international safety standards. Given the safety performance and time for the implementation, it may happen that during the period between two periodic safety reviews, only the design started over change. And we already include in the new periodic safety review report as if the change has already been implemented to show what will be for the next 10-year period. So also the extent to which the safety documentation including the licensing basis remains valid. These are what we have to analyze. Now how do we do that? How do we do that? This special safety guide specifies so-called safety factors relating to the plant. Here we had in the original safety analysis report, I don't know, one to 19 chapters describing each chapter its specific topic. Here we have safety factors analyzed. The first safety factor, safety factor one is the plant design. To determine the adequacy of the design of the nuclear power plant and its documentation by assessment against the current licensing basis. And national and international standards, requirements and practices, yes. So the current design to the current requirement. Then safety factor two, actual condition of structures, systems, and components as well as the SSTs important to safety. So this will contain the assessment to determine the actual condition of SSTs important to safety and so to consider whether they are capable and adequate to meet design requirements. At least until the next periodic safety review. In addition, the review should verify that the condition of the SSTs important to safety is properly documented as well as reviewing the ongoing maintenance, surveillance, and in-service inspection programs as applicable. So what do I analyze when I'm doing the in-service inspection of a pipeline? The third one measure, well, the residual wall sickness. So we are measuring the residual wall sickness because we are trying to identify those cracks and the growth of the cracks inside the wall which will then determine the decrease of the effective wall sickness of the pipe. Now of course, there will be a critical wall sickness below which we will not withstand the pressure inside. So I am analyzing the speed of the growth of the crack. And I'm predicting for the next 10 years I have to justify that I still will have a sufficient margin in the wall thickness in order to operate without any problems. So this is what the in-service inspection is usually doing at the nuclear power plant. So in other words, the aging of the pipeline is measured and predicted for the next period. And this is the actual condition of structures and systems and components. For active components, the trends of the failure rates, how often they are failing or vibration in a pump, increase of vibration in the pump and the region, the pressure region where this vibration is going on, how the pump would behave in the real actuation when it was needed, will it perform, will it still perform its function or it will fail. So these are the investigations which I have to include in the justification of compliance with the requirement. Equipment qualification. Now we had a lecture about equipment qualification. We have to reassess the equipment qualification in order to determine whether the plant equipment important to safety has been properly qualified according to some new well knowledge because as Alex said, the equipment qualification is not just a one-shot project. We have to revisit and reconsider the equipment qualification. So including the environmental conditions and whether this qualification is being maintained through an adequate program of maintenance, inspection and testing that provides confidence in the delivery of the safety functions until at least the next periodic safety review. Then safety factor four is the aging. You can see that these safety factors are somehow interrelated. So there is some cross-cutting areas where I am analyzing something at one safety factor it comes back again in the other like aging to determine whether aging effects affecting the SSC is important to safety are being effectively managed and whether an effective aging management program is in place so that all required safety functions be delivered, it is repeated many times. Deterministic safety analysis. We have to determine whether the existing safety, deterministic safety analysis are complete and are still valid or I have to do something, I have to redo some of them or all of them depending on all kinds of design changes, organizational changes, whatever. So all the changes has to be considered and also I have to consider the new developments in the analysis techniques. So earlier we had how it was called the original deterministic code which the original design took into account, it was thermo-dynamica, something, I don't know. Today there is a much more developed program and also we used relapse and other deterministic codes which are using more precise or more verified correlations and so on and so on. So definitely for those codes, using those codes the conclusions from the analysis may change drastically whereas we thought that, okay, the temperature, the cladding temperature will go up to 800 degrees during a process. The new calculations may show that at 400 degrees the temperature starts decreasing and so on and so on. We can see easily such changes in the analysis results. So probabilistic safety assessment, to determine to what extent the existing PSA is complete and remains valid, again the same. If I'm changing something in the design, in the system, I'm making some kind of physical separation of pipelines which were not there before then definitely the assumptions I did during the probabilistic safety analysis will change because I can be sure that the structure of that pipeline will not affect the other pipeline because there is that thick concrete wall between the two which did not exist at the beginning. Hazard analysis is also the 7th safety factor. Yes, Chapter 15, Chapter 15 of the final safety analysis report, yes. So we'll have to review all those safety, deterministic safety analysis which are included in the safety analysis report, the final safety analysis report. We have to update the final safety analysis report according to the changes what we discovered. Okay, I'm talking about Chapter 15 according to the US terminology which is coming from but it may be different to the IAEA terminology I don't remember. So the deterministic safety analysis or the safety analysis as a whole including the deterministic and probabilistic safety analysis should be reviewed. Okay, so if the safety analysis is in Chapter 5 or 6, then I have to, yes, just. Someone from the Chapter 15, okay. Talking about the reactor, the floor and talking about. Oh, I see, okay, yes, yes. They may be included. Yes. Are included in Chapter 15 and can. But they must be included in, so there is a reference between, there is a reference from Chapter 5 or 6 to the Chapter 15, but the basis for that is the Chapter 15. The safety analysis chapter is the one which, yeah. In any case, the final safety analysis review is updated with the information included in the periodic safety review, okay. So I will have a new final safety analysis, the final safety analysis report is a leaving document together with the plant safety. So it is not a constant document which was created at the beginning of the lifetime of the plant when we issued the operating license, but it will be changing by those changes which are in the periodic safety review, okay. Hazard analysis. The hazard analysis is also to be reviewed because there are also changes in the resistance against some hazards. We did awful lot of work to separate the original cables which were installed at the plant. We had at the beginning cable trays with that bunch of cables and nobody knew what kind of cables were there. And then we wanted to do the risk, a fire risk study. It was almost impossible. So we decided to first put everything in order, which means separate the safety train cables from all the rest, which was easier to disconnect those cables which were there and rebuild new cable routing, okay. After that, we already knew that the physical separation of those cables are there and we can do the fire hazard study, internal fire hazard study. So this kind of analysis should be also included in the periodic safety review in order to show either the relevance of the old study or to analyze the new situation. Then safety performance. It is to determine whether the plant's safety performance indicators and records of operating experience, including the evaluation of root causes of plant events, indicate any need for safety improvements. Do you know what are the safety performance indicators? Do you have it in your plants? Safety performance indicators? Safety performance indicators are those number or numerical values which characterize some part of the safety. So a performance indicator system would somehow characterize the behavior of the plant's safety in terms of numbers. So it is for the managers to show that the number of reactor scrams in the last 10 years are remained below one or something like that or the radioactive effluents were that much and that much in the last year. So these are yearly evaluated or quarterly even and these performance indicators can indicate some deficiencies, some trends, some changes and so on and so on. Therefore the safety performance in terms of performance indicators should be investigated and should be reported in the periodic safety review report. Some regulatory bodies require the usage of a plant's safety performance indicator system and in some countries the reporting of the safety performance indicators quarterly or yearly is required by the regulatory body. So to be able to follow the plant safety performance. Then of course the records of operating experience and the root causes of plant events depending on what kind of events are to be analyzed by root cause analysis. Not all the events of the plant are analyzed by root cause analysis but there are some where I want to know what was the root cause of that particular occurrence. I have to perform root cause analysis and the root cause analysis will indicate, will show me what was the root cause for that particular occurrence. It is not easy, let me say so. It is not easy because some occurrence may have potentially different causes. So what has been the root cause in the particular case? If we are closer in time to the event we are very likely able to identify what was the root cause. Otherwise later, a year later or two years later just you want to analyze what was the root cause. You are losing already so much information that you can already approach with a probabilistic approach the causes. And you can already just identify the contributors to the events of the different causes. So the root causes will give us lots of interesting information and it may indicate the need for some safety improvements. Then the use of experience from other plants and research findings. Other plants and research findings is something which needs the data investigation or information collection from sources which contains the international experience, operating experience. VANO, the World Association of Nuclear Operators, runs such system which collects the operating experience in terms of incident reporting, also event reporting, then the performance indicators and so on and so on. And all the members of VANO will be able to access this information. The IAEA is running the IRS system, incident reporting system, which will also be a source for such international experience. And there are some regional so-called information collection systems like system of some owners group members or more broad international systems. So there are lots of ways of collecting this information from. And the plant operator is part of some of these organizations, therefore they will be accessed to this information. And they have to use this information in order to see or to determine whether there is adequate feedback of relevant experience from other nuclear power plants and from the findings of research and whether this is used to introduce reasonable and practical safety improvements at the plant or in the operating organization. So if I see that the other nuclear power plant, somewhere in another country, similar nuclear power plant, discovers some safety deficiency and it comes into the operating experience collection system. I also have to take care about this information. I also have to investigate whether it is relevant to my plant. And if it is relevant to my plant, I have to use it. I have to use this. So this is the feedback of the operating experience from other operating organizations. Then safety factors, organization, management system, and safety culture. This is something soft topic, let's say so. Usually not treated very seriously, however, the organizational factors, the management systems, and the safety culture, evaluation of the safety culture at the plant is very important. Unfortunately, the assessment of these safety factors is based on some kind of self-assessment, some self-assessment techniques, which means I am analyzing, I am assessing myself, my performance, and trying to find out whether I did something wrong. It is very difficult to acknowledge that I did something wrong. And especially when it goes to some periodic safety review report, which goes to the regulator. So I'm still a little bit skeptic about the truth of the content of this safety factor in the different periodic safety reviews. In an open environment where the management and the people are open to, and they acknowledge that they can do mistakes, then the acknowledgement of being able to do mistakes, this is important. In such an environment, it is easier. But in other cases, when it may have some bad consequences, bad personal or organizational consequences, it is difficult. So it is to determine whether the organization, management system, and safety culture are adequate and effective for ensuring the safe operation of the nuclear power plant. Now comes the question. If there was some bad event at the nuclear power plant, which clearly shows some evidence that it was due to some organizational factors, then this should appear here in this chapter or under this safety factor. So that is really difficult. Then comes the procedures to determine whether the operating organization processes for managing, implementing, and adhering to operating and working procedures and for maintaining compliance with the operational limits and conditions and regulatory requirements are adequate and effective to ensure the safety. So procedures and organizational processes. These are those processes which are regulated by the procedures in the maintenance, in operation, and other areas. So these are administrative procedures, and I have to assess those whether they are serving to help keeping the safety on the required level or not, or there is some deficiency. Sometimes it happens that such audit will indicate some deficiencies which affects the safety. Therefore, those results of those audits should go into this safety factor. And human factors. This is to evaluate the various human factors that may affect the safe operation of the nuclear power plant and to seek to identify improvements that are reasonable and practicable. Human factors we were talking about and we talked about the human factor engineering requirements and elements. All this should be discussed here. Then factors relating to management, emergency planning. It is to determine whether the operating organization has in place adequate plans, staff, facilities, and equipment for dealing with emergencies. And whether the operating organization's arrangements have been adequately coordinated with the arrangements of local and national authorities and are regularly exercised. Such emergency drills, let's call it like that, are time by time organized on the site. And time by time in cooperation with the local authorities, they are organizing such emergency drills in the region which may be affected by the nuclear power plant. And the results of these drills and the description of what are those facilities, equipment and staff and so on should be demonstrated that they are in compliance with the actual regulations. So this is an interesting topic again because people usually don't like to participate in such drills, especially if you have nothing to do with the nuclear power plant. So I remember when we lived in Pax, they distributed these gas masks for the children. Not for the adults, but for the children. In case something happens, and my little kids were so happy, they received a new toy. So they wanted to try it all the time. But then we said, okay, guys, this is not for playing, we took them away and put on the shelf just for the case. But there were some drills where, yes, they had to use it, so yes. So we were not that happy, but they took it as an interesting game. Safety factors relating to the environment. Now the radiological impact of the facility to the environment should be discussed to determine whether the operating organization has an adequate and effective program for monitoring the radiological impact of the plant to the environment, which ensures that the emissions are properly controlled and are as low as reasonable achievable, ALLARA. So you know what ALLARA means, so everybody knows what ALLARA means, I think. So during normal operation, the releases should be ALLARA. So that's it, that's the message, and you have to justify it, you have to show it. So basically in our plant they can demonstrate that the releases from the Paxchia nuclear power plant is lower than the radioactive releases of the conventional power plant near our settlement where we are living, which is funny but true. So for some reason the radioactive releases of that traditional power plant in normal operation is almost one order of magnitude higher than the gas releases of the nuclear power plant, which is a good news to the nuclear power plant, but not a good news for the people who are living though it is well below the limits, of course, so it is not bothering anybody, but the measurements show such difference, okay. So basically this is the 14 so-called safety factors to be analyzed. Now there is one more important issue which is the review of the physical security of the nuclear power plant. It is generally not included in the periodic safety review. Physical security is something very sensitive, so it should be kept in secret and so on and so on. Of course there should be a review, periodic review and with the new developments definitely the security should increase, but it is not generally included in the PSR because of the sensitivity of the subject and the need to ensure confidentiality. This should be reviewed periodically by the appropriate national authorities, so there are dedicated authorities who will do this review. However, some operating organizations may decide to review physical security as a separate safety factor in the periodic safety review. We could see in some countries in the periodic safety review discussion of the review of the physical security or the physical protection. Okay, now the PSR findings, definitely everybody would like to see, so why do we do this review? To formulate the findings, to formulate the issues and of course for political reason we are formulating first the positive findings that this has improved, this has been better and this become more powerful and so on and so on. So that is the strength of which we are indicated, where current practice is equivalent to good practices as established in the current codes and standards and so on. But the negative findings, which are the deviations where the current practices are not of a standard equivalent to current codes and standards or industry practices or do not meet the current licensing basis or inconsistent with operational documentation of the plant or operating procedures. So there are three types of deviations in the first for which no reasonable and practicable improvements can be identified. There are such design deficiencies which simply I cannot do anything or some other problems which I cannot do anything. Whatever I invest in it will remain. Therefore, this deviation has to be indicated and later it will be decided by the regulator whether they can live with it or not. If not, they may even obliged to shut down the plant. But in most of the cases under some conditions they allow continued operation. Then there are deviations for which identified improvements are not considered necessary. Well, it's interesting to have such deviations which you don't have to care about. And deviations for which safety improvements are considered necessary. Now those deviations will be a question of licensing and a licensing issue. Then the regulator will tell you that within the next five years period you have to solve the problem and declare the problem be solved. Okay. And these are the periodic safety review tasks. I will skip it. You know, sorry. There are four points which to be mentioned. Preparation, this is a long time process because you have to form a group of people. You have to form a project which will prepare the periodic safety review report. Also, which will perform those necessary safety analysis tasks which are included in the periodic safety review. Conduct of the periodic safety review when you are executing it. This is also a long time process. All together, actually, the preparation and the execution can be more than two years according to our experience. And the regulatory review of the findings with the regulatory decisions. And then comes the finalization of the integrated implementation plan where you want to get rid of the negative findings. You are describing the implementation plan, how to proceed in the next two, three, five years period in order to improve the safety or in order to be on the level required at this time. Okay. So basically, this was the periodic safety review presentation.