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In spite of the fact that the VVER has been present on the market for 40 years ago it is still subjected to upgrading in order to improve the economic parameters of the fuel cycle and increase the fuel reliability during its operation. In the process of VVER gaining operation experience 1 , numerous advancements of fuel assembly FA design characteristics have been realized 2. Achievement of these advancements required the constant upgrading of existing FA designs to increase their service life, improve their operational reliability and reactor safety and economics of fuel cycle Table 1.

A big effort has been made in the reactor community to reach these goals. Below we show the effect of these advancements and modifications on the neutronics and operational characteristics for the initial and end of the modernization stages Table 2 and Figure Now it is clear that improvement strategies refueling and development of the neutronic codes helped to optimize the core load. This allowed to increase the fuel cycle economy and to substantiate the overload in-in-out type scheme which provides to lower neutron leakage from the core and to decrease the radiation load on the reactor vessel.

Advancement of the FA base design allowed to significantly improve fuel efficiency by: - using of zirconium spacer grids and reducing of FA shroud tube thickness resulting to the increasing FA multiplication factor; - grading of fuel enrichment in the cross-sectional plane of FA thus reducing a peaking factor of linear fuel pin powers and ensuring by this a lightened realization of core loads with reduced neutron leakage; - increasing of initial enrichment of fuel feed and reducing of fresh fuel replaced during fuel reloading; - using of burnable gadolinium Gd2O3 poison integrated with fuel ensuring desirable fuel load arrangement and leading to improvement of load safety; - raising the average fuel enrichment up to 4.

Presented results are contribution to the gradual evolution of VVER fuel assembly. Selected alternative variants of? Influence of in-assembly power peaking on operation limit margins is evaluated. Khalimonchuk2 , A. Kuchin2 , 1 : T? The paper discusses a probable significant global power field tilt caused by a random small change in multiplying characteristics of fuel assemblies in a physically large reactor. The reaching of normal power noticeably decreases the power field tilt, at that. Comparison is made between neutronics characteristics of Uranium-Gadolinium and Uranium-Erbium fuel cycles.

The study shows that use of Erbium as burnable poison allows decreasing the peaking factor in the core. Meanwhile residual Erbium at the end of the fuel cycle makes it necessary to increase fuel enrichment. This paper discusses the methodology and the performance of this tool. The first method is the imitation of the equilibrium pattern that is the optimized pattern in case the cycle length and the operation conditions are constant and the same shuffling pattern is repeated from cycle to cycle. The imitation utilizes heuristic rules but is also stochastic in nature.

The function of the imitation algorithm is to provide initial patterns quickly for the next optimization phase, which is performed either with the stochastic guided binary search algorithm or the deterministic burnup kernel method depending on the choice of the user. The former is a modified version of the standard binary search. The search is guided so that the swapping of fuel assemblies always starts at or near the most restrictive assembly. The kernel method creates burnup kernel functions to estimate burnup variations that are required to achieve desired changes in the power distribution of the reactor.

Nuclear safety studies | AERB - Atomic Energy Regulatory Board

Although, the tool was especially designed for Loviisa reactors the basic ideas of the search algorithms could be utilized in any VVERreactors. Furthermore, the tool has been used to search loading patterns for upcoming cycles of Loviisa plants with good results. Depending on the strictness of the safety margins, finding a feasible loading pattern takes from few minutes to couple of hours. The quality of the found patterns is equal or better compared to the manually designed loading patterns. The code was presented last year and has already been used for VVER type nuclear reactor fuel loading pattern of optimization but has more general abilities.

Optimization has allowed us not only to find long cycles, but also to evaluate the effectiveness of new fuel types and to estimate options and feasibility of long cycles and number of necessary fresh fuel assemblies. Assembly power distribution Kqi and that of separate sections of assembly Kvi , j are compared. The paper shows the different densities of distribution of calculated vs. Since it? The system is being adapted according the utility needs. With its flexible and modular framework successfully responses to the plant operating needs and advances in nuclear fuel cycle strategies and fuel design.

Modular framework allows for easy modifications of the system and implementation of new methods in physical modules. These facts have been confirmed by successful upgrades and more than 13 years of reliable operation of the core monitoring system in two countries and six VVER units. Method of accurate determination of VVER reactor thermal power applying ex-core detector signals.

A coolant flow through the reactor is usually determined after annual outages at Slovak NPP in two distinct ways. First method is determination on the basis of the secondary system parameters — measurement of thermal balances. The value achieved by this method is used as the input parameter in the Table of allowed reactor operation modes. The second method draws from the primary system parameters — measurement of primary system hydraulic characteristics.

Flow nozzles used for the measurement of feed water flow behind high pressure heaters were replaced at both Bohunice Units during outages in The feed water flow behind high pressure heaters is one of the main parameters used for the determination of coolant flow through the reactor by the first method. Compared to the measurement executed during previous fuel cycles, the calculated coolant flow through the reactor decreased considerably after the change of flow nozzles.

This change was not proved by the parallel measurement of primary system hydraulic characteristics. Later it was found out that the original flow nozzles used for 25 years were substantially deposited original inner diameter of the nozzles was reduced by about 0,6? Therefore feed water flow measurement was untrustworthy within the recent years. On the findings stated above, Bohunice NPP has decided to increase coolant flow through the reactor by changing the reactor coolant pumps impeller wheels.

Modification of impellers wheels is planned within years to The 4 impeller wheels at Unit 4 and 2 at Unit 3 were replaced during the outage in In general a reactor is initially started up from a precondition stage ? C by withdrawing control rods or by changing the boron acid concentration in primary circuit coolant until the reactor is slightly supercritical, thus producing an exponentially increasing neutron population on a very long period. As the neutron population increases, the fission heating and thus the reactor temperature increases. This increase in temperature produces a decrease in reactivity.

That would lead to the slowdown of increasing the neutron population and the reactor power should saturate at certain higher level. Generally, core loading patterns are designed in such way that isothermal reactivity coefficient should already be negative at the Cold Zero Power CZP. To meet this requirement could be problematic, particularly for the first core loading, when all fuel assemblies are fresh.

Due to the first criticality start-up of the Nuclear Power Plant NPP Mochovce units 3 and 4 in the near future, detailed analyses of core parameters are required by the Slovak Regulatory Authority to support safe operation of the nuclear facility. The article introduces determination of the thermal reactivity coefficients, especially summary isothermal and moderator density reactivity coefficients between ?

C and ? C with step of 2? The work presents calculated critical parameters, especially critical boron acid concentrations at given coolant temperatures and position of the 6 th control assembly group.

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Numerical iteration procedure was applied to calculate the critical parameters as a substitute for a critical experiment. Geometrical and material models were created in compliance with the reactor design and the first fuel loading of the NPP Mochovce 3 and 4. All mentioned calculations were performed by computational code MCNP5 1. In design justification, considering a conservative approach, the lifetime of absorber element AE is limited to 10 years, among them not more than 3 years as a part in the control group.

The paper reports on Moscow final meeting of working group H? Elaboration of methodology of calculating the core design engineering factors? Kurchatov Institute? Participants of working group?


Name-list of the participants and titles of their presentations may be found in the paper, which also gives brief information on the designation of presentations. The macrocode qualification is key point of the redefinition of the engineering factor methodological component. A hyperbolic dependence of the methodological component on the power level is not strict requirement and from the point of some safety criteria, e. In the paper is presented development of the methodological component of volume power distribution engineering factor F0 like the prediction interval and trapezoidal width of confidence of linear regression including simultaneous interval Scheff?

There were determined standard deviations calculated versus measured power distributions like time and spatially varying uncertainty. In reference to the above-mentioned standard deviations the true values of power distribution were determined according the Bayes? In this sense there were realized alternative evaluations of the methodological components of engineering factor F0. In the design of VVER core loadings, the important parameters that characterise reactor safety are the effectiveness of emergency protection and the worth of control rods.

The calculation uncertainties of those functionals are also important for the design. The paper discusses the methods of accounting the uncertainties by comparing calculated and measured data. The paper proves that the deviations between BIPR -7 calculated data and measured data are distributed in accordance with close to normal law, while the parameters of distribution such as mean and standard deviations , which were determined for the available comparison samples, are practically identical to those determined for the applicable general samples.

Systematical deviation in average absolute differences between calculated and measured TC temperature rises for the fuel assemblies 1 st year in the active zone was found. The influence of the cleaning of the reactor shaft bottom and the throttling orifices on the methodological? Nowadays, there is a tendency to use best estimate plus uncertainty methods in the field of nuclear energy. This implies the application of best estimate code systems and the determination of the corresponding uncertainties.

In this paper the uncertainties of the neutronic calculations at core level — originating from the uncertainties of the basic nuclear data — are presented. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution and the rod worth are shown.

After that the preliminary evaluation of the uncertainties of the neutron kinetic calculations are presented for a rod movement transient at HZP state, where the uncertainties of the time dependent core and assembly powers and the dynamic reactivity were evaluated.

In both cases, we will see that the most important quantities — at core level and at HZP state have a considerable uncertainty which is originating from the uncertainties of the basic cross section library in these investigations. In order to perform this large task, 3 phases were defined in the benchmark and these phases include the survey of the uncertainties of the stand alone neutronics Phase I , the time dependent neutronics, stand alone thermo-hydraulics, the fuel behavior calculations Phase II and the system phase, as well.

In this paper we concentrate on the determination of uncertainties of? Phase I, Exercise I-3 and partially on the determination of uncertainties of the? Time-Dependent Neutronics? It is important to note that only the uncertainties of the. In the paper is represented the method of statistical account of uncertainties to obtain estimation of DNBR and value of heat flux per rods length unit.

The method is based on the special developed FA model which includes mathematical models of energy field calculations, FA thermomechanical deformations, thermohydraulics and technological processes. FA thermohydraulics calculations carry out using cell by cell thermohydraulic code SC-1 1. In these calculations were taken into account deviations in different parameters Statistical properties of DNBR minimum value in FA and in specified rods, statistical properties of heat flux from the surface of the fuel rods and coolant heating in specified cells is determined using statistical simulation of large number of random parameters affecting this values.

It have been estimated the value of reducing factor for DNBR and this value was analyzed for sensitivity to different factors: FA power, relative axial heat flux profiles, rods energy distribution, the law of statistical distribution random parameters, the number of? Also the estimation of engineering safety factors for coolant heating and local heat flux have been obtained. The validation of depletion codes is an important task in spent fuel management, especially for burnup credit application in criticality safety analysis of spent fuel facilities.

The fuel assembly has initial enrichment 4. Criticality calculation for pools at reactor in Slovakia Vladim? Severe accident? The Fukushima accident has triggered urgent discussions about the need for such type of investigation. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization.

For some time discussion was going on, whether the in-vessel retention can be applied for the VVER type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity.

In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The scaling ratio for reactor surface is but the scale of the elevation is to ensure correct representation of the natural circulation processes. The heat flux is generated by electric heaters. A large number of pressure, temperature, water level, void and flow measurements allow tracking of the coolant behavior and surface temperature. In the experiments, the area of the coolant channel can be adjusted in the critical section where the gap size is minimal.

This benchmark was defined by A. Kotsarev, M. Lizorkin and R. It is focused on investigation of transient behavior in a VVER nuclear power plant. Its initiating event is the opening of the main isolation valve and re-connection of the loop with its main circulation pump in operation. A water slug with lower temperature of oC enters the core which leads to a reactor power excursion. The activation of the reactor scram is caused by the corresponding power level signal. As a result of the scram, the turbines are turned-off by closure of the turbine isolation valves. All main circulation pumps remain in operation.

Each participant had to use its own developed input plant data deck and parameterized library of macroscopic cross sections. The initial reactor power tuning and burnup calculations were performed with the code DYN3D by both participants. The second part shows the comparison of important global and local time dependent parameters. The trends of main calculated parameters are very close. The calculated nuclear and thermalhydraulic phenomena like reactor power, cold leg mass flow rates, cold and hot leg coolant temperatures, upper plenum pressure, pressurizer water level and steam generator powers agree quite well.

The differences in the time behavior of compared parameters are discussed at the end of paper. From among the advanced new features, the possibility of using triangular geometry and the automated cutting the triangles into further ones at many levels was utilized for an attempt to reach a converged solution of the Second AER Kinetic Benchmark Problem.

Most of them can be classified into the group of the? Although the most important? An advantage of these characteristics must also mentioned, namely this approach is representing the real situation concerning the modeling uncertainties as far as the variance of the different solutions is regarded. The first and second kinetic benchmarks 1 are exceptions from this point of view because their definition makes it theoretically possible to have the converged reference solution. In this sense, the two latter mentioned benchmarks can be regarded as?

Nevertheless, the efforts for determining their converged reference solutions seem not fully satisfactory, especially regarding the AER Benchmark Book? A very important, large step in this direction was made by N. Kolev, R. Lenain, and C. An extrapolated to zero mesh size reference solution was recommended for the steady states and it revealed significant deviations from the solutions obtained from the nodal codes. The significant deviation raised certain doubts in some participants concerning the advised reference solution. The goal of the present work is to contribute to the existence of the reference solution of the second kinetic rod ejection benchmark in the AER Benchmark Book.

In the present paper the possibility of performing a multiscale modeling of the hydraulics of a reactor pressure vessel RPV using the system code ATHLET 1 is considered. The main focus of the work is the modeling of the active core. There are several length scales that can be used in the modeling of RPV , namely 1 macroscale, or the set of fuel assemblies sectors ; 2 mesoscale or an individual fuel assembly; and 3 microscale or a pinwise modeling of one or several fuel assemblies.

In all above mentioned cases the modeling structures are represented by the systems of parallel channels, hydraulically coupled with each other in the transversal direction in the absence of the walls. The modeling allows using both meso- and microstructures and the object nodalization of the upper and lower part of the reactor uses macro- and mesostructures.

To improve the modeling of the flow in the transverse direction the convection term in the ATHLET momentum equation is corrected to give the possibility to simulate two- and three-dimensional flows. A diagnostics of initiating events, states of reactor systems and relevant safety variables is important for safe operation of an NPP. On the one side one needs to correctly interpret signals from measuring instrumentation.

On the other side a fast classification of the transients significantly improves the reliability of NPP operation. This paper presents an original approach for detection of NPP equipment failures and for classification of complex transient states of NPP. It is based on the use of artificial neural networks.

The sequence and the principles of network-learning, as well as the justification for selecting the type of neural network are presented. The paper contains two applications of the approach: a the integral and local estimates of recognition accuracy of latent and false rejections in measuring channels and recommendations on the application of the implemented methodology in automatic process control systems of NPP units; b a robust classification of the transients in German NPPs using the implementation of the ALADDIN tool developed in the Halden Reactor Project.

The module of an estimation of parameters of a preemergency condition representative fuel rod contains a necessary set of correlations for their conservative estimation. It is necessary to underline, that this module is under construction by a principle of the minimum sufficiency for the emergency analysis, and for reception of more realistic estimation, the given module can be replaced with the interface of the resource fuel rod code. The description of total modules is presented. Heat transfer in the gas gap of a fuel rod affects the behaviour of the reactor core also in the scale of the whole reactor.

In the VTT? Typically burnup is taken into account in a rather simple way, e. Implementation of a detailed fuel model in the reactor dynamics codes would not be a suf? As the properties of a fuel rod change remarkably during its normal reactor life, the initial state of the transient must be somehow determined.


In this paper, the statistical version of the fuel performance code ENIGMA has been used to determine the properties of a representative VVER fuel rod at various points of its reactor life. The aim was to? Calculation results are compared with measurements from the? The calculations are performed with the reactor dynamic nodal code DYN3D.

The usual node-averaged approach is applied in the first step. The BSS Directive tackles nuclear industry as well as other practices or situations where people or the environment could be affected by ionising radiation. The directive forms a harmonised system related to nuclear and radiation safety in the EU together with two other directives, namely:. The BSS Directive is a comprehensive document repealing altogether five other legal documents. It is a very technical document using more than physical parameters.

It introduces a set of new concepts as well as new or updated physical parameters, such as a new dose limit for a lens of the eye and so-called clearance levels used for solid material released from the regulatory control. Among important concept is requirement to use current state of technical knowledge in the justification process. In addition, a control over building materials as well as natural sources is strengthened. The analysis shows for example that the licensing of a facility is strengthened, control over discharges is introduced and dilution of radioactive materials is discussed enabling re-evaluation of decommissioning plans of nuclear facilities.

As a part of post-Fukushima activities emergency preparedness in a case of nuclear or radiological accident is strengthened in the BSS Directive. Challenges of the MS and particularly of nuclear industry when implementing the requirements are discussed in details. The main target group of information activity are schoolchildren with their teachers.

Most of them are from the 8th and 9th grade of elementary school, age 14 to Every year some youngsters visit the Information Centre. The visit consists of a live lecture about nuclear technology followed by the demonstration of radioactivity and a guided tour of a permanent exhibition. Since we monitor the opinion trends by polling about youngsters every year. The youngsters are polled before they listen to the lecture or visit the exhibition in order to obtain their opinion based on the knowledge from everyday life. In the paper we will present, summarize and comment the trends over the last 21 years.

In we upgraded some of the existing and introduced some new exercises. The pulse mode operation exercise was upgraded by installation of new data acquisition system and development of new graphical user interface GUI by using LabVIEW software. The critical experiment exercise was upgraded by adding a new detector.

Now we monitor neutron population with two independent fission chambers on different locations. In the past the void reactivity coefficient exercise was performed by inserting Al tube into various positions in the reactor core and measuring the corresponding reactivity changes. In order to make the exercise more realistic, we installed a pneumatic system for generating air bubbles just below the core.

Analysis of Reactor Fuel Rod Behavior

The system consists of a system of valves, flow meters and Al tubes for conveying air under the core. The trainee can adjust the air pressure proportional to the flow rate and the location in the core at which the air bubbles are generated. The flow rate at individual locations is measured. The aim of the exercise is to measure reactivity changes versus flow rate and air bubble position.

The second new exercise was measurement of water activation. In this exercise we installed special system which pumps the water through the core at a constant flow rate to the reactor platform, where the water activity is measured with a portable GM tube and two spectrometers, a semiconducting HPGe and a scintillating LaBr. The purpose of the exercise is to measure the 16N and 19O gamma line intensity and dose rate versus reactor power. It can be seen that the relationship is linear. The third new exercise, named in core flux mapping, was performed by measuring the axial fission rate distribution at various radial positions in the core.

We used CEA — developed mini fission chambers and a special home developed system for moving the fission chamber in axial direction and measuring the count rate versus FC position. In the paper we present the new exercises in more details, first results and plan for the future. Severe accidents - Accurate calculation of the iodine behaviour in the containment is of crucial importance in determining the potential radioactive source term to the environment under light water reactor severe accident conditions.

Of particular importance is the behaviour of iodine in the gas phase, particularly organic iodine which is difficult to remove by filtration e. The iodine behaviour is closely linked with the containment thermal hydraulics that have a major influence on the distribution of iodine throughout the containment atmosphere and sump. In the EC SARNET2 network of excellence, the predictive ability of current severe accident codes in this important area was assessed through two benchmark exercises. In the first, the codes were assessed against data from the German THAI Iod and Iod integral tests, in which particular attention was paid to molecular iodine transport with atmospheric flows, and the iodine interactions with steel surfaces.

Thermal hydraulic conditions in the containment were simpler, while realistic fission product sources were used and radiolytic interactions of iodine, e. The two benchmarks are thus complementary. In the FPT3 exercise the calculations could predict fairly well the general, well-mixed, thermal hydraulic conditions. For THAI, where there are more detailed measurements, significant differences were noted for atmospheric flows and relative humidities.

Substantial user effects were noted in both exercises, indicating the need for improved user training in phenomenology and optimum code use. Alexander D. Vasiliev 1 , Juri Stuckert 2. Tulskaya, Moscow, Russian Federation 1. Karlsruhe Institute of Technology, P. Box , Karlsruhe, Germany 2.

The test bundle was made up of 21 fuel rod simulators which are placed in the square set. Heating was carried out electrically using tantalum heaters. The rod claddings were identical to that used in PWRs. The bundle was electrically heated in steam from K to K with the heat-up rate of 5. A steam explosion, in the frame of nuclear reactor safety, is an energetic fuel coolant interaction process, which may occur when the hot reactor core melt comes into contact with the coolant water. Steam explosions are an important nuclear safety issue because they can potentially jeopardize the primary system and the containment integrity of the nuclear power plant.

If metallic zirconium is present in the corium melt the oxidation of zirconium may significantly influence the fuel coolant interaction process, as observed in experiments. To find out the reasons for this qualitatively different behaviour, the experimental results were investigated in comparison and computer simulations were performed. In the paper, the performed analysis of the influence of the zirconium oxidation on the steam explosion energetics will be presented and discussed.

Based on the experimental findings, the hydrogen film hypothesis is proposed, claiming that only a limiting amount of zirconium may be oxidized during the premixing phase in subcooled conditions and that the remaining amount of unoxidized zirconium is available for the oxidation in the explosion phase. Various computer simulations were performed with the MC3D code to support the hypothesis and to get additional insight.

It may be concluded that the proposed hypothesis reasonably well explains the observed experimental differences. Suggestions for further analytical and experimental work will also be given. In Safety Analysis Report SAR Chapter 6 it is demonstrated that global hydrogen concentration stays below limit using one compartment containment model. It is assumed that due to well connected compartments and large volumes possibility for appearance of high hydrogen local concentrations is very low.

It is still possible to see some temporary local increase in hydrogen concentration due to non uniform hydrogen production and hydrogen stratification, especially for Beyond Design Basis Accident BDBA conditions. In both situations PARs are able to prevent situations where combustible gases can reach flammable concentrations. The experiment resulted in channel voiding as a result of sodium boiling and a clad melting. Only some fuel melting took place. The main attention during validation phase was given to modelling of fuel-clad heat transfer, pin mechanical behaviour creep deformation of the clad, account for direct fuel-clad contact, fuel swelling, and release of fission gases within the fuel , movement of molten materials.

Nuclear energy, an energy for the future. Christophe Behar. France nuclear fuel cycle strategy is driven by the conviction that nuclear energy is to be maintained as a stable pillar of the French energy mix in the long term. Fukushima accident has triggered deep and comprehensive safety and economic assessments which have reinforced the fundamentals of this strategy.

Thus sustainability is a major driver of France nuclear strategy, among which fuel cycle policy plays a major role. It is based on the recycling of spent fuel in order to save energy resources and to offer a better management of nuclear wastes. On the other hand France has also a large feedback experience at industrial size with sodium cooled fast reactor, indispensable tools to burn efficiently the plutonium coming out from light water reactor Mox spent fuel or sodium fast reactor Mox spent fuel.

Owing to this experience in La Hague, Phenix and Superphenix, France is developing Generation IV fast reactor along with their related fuel cycle, these systems being the keystone of sustainable energy nuclear development. In the global context of the increasing demand for energy, while fossil fuel resources gradually reach exhaustion, energy management, as a vital need and a factor of economic growth, is a major challenge for the world of tomorrow, facing with the desire to reduce greenhouse gas emissions.

In the worldwide demand for energy will have doubled under the combined effects of population growth — to 9 or 10 billion by — and the development of emerging countries. At a time when resources are becoming scarce and it is increasingly urgent to combat climate change, it has become indispensable to ensure competitive low-carbon energy sources. The advantages of nuclear energy — electricity production without greenhouse gas emissions — make it a promising solution. Although nuclear energy has significant advantages in this regard, it faces safety, resource sustainability, and waste management issues that must be met through continuing technological innovation, to improve its capability to continue satisfying the long-term demand for energy, which is increasing endlessly.

FNRs appear to show the best potential to reach 4 th generation criteria for industrial deployment in the middle of the 21 st century, or even earlier if needed, based on the accumulated operating experience of more than reactor. The Astrid reactor, a MWe prototype reactor, is the indispensable step before an industrial deployment. It is representative of the main necessary industrial characteristics and its demonstration capabilities are designed for the qualification of innovative concepts.

Multiphysics - Universitat Politecnica de Catalunya, C. Jordi Girona, 31, Barcelona, Spain 2. The low frequency noise is a phenomenon observed in some PWR reactors causing an influence on the normal operational behavior of the power plant. Fluctuations in the neutron flux density, in the low frequency range up to 4Hz, generate noise in the neutron instrumentation incore and excore neutron detectors that could affect to the limitation and protection system of the reactor core, even it can activate a scram signal due to high neutron flux.

In this project has been carried out a deep analysis of the behaviour of a PWR reactor core and the signal of the neutronic instrumentation when a set of inlet thermalhydraulic perturbations are applied. The cross sections have been obtained using the SIMTAB methodology and the signal of incore and excore neutron detectors are directly extracted from PARCS through a previously validated modifications incorporated in the code. Thorium is very attractive as sustainable resource in nuclear industry for its large resource inventory, high potential conversion ratio in thermal reactors because of the produced U, and less amount of long life MA products.

In a series of previous works, the feasibility of maximizing thorium utilization and minimizing the refueling effort of uranium fissile under the framework of HTR-PM has been investigated. In this work, further investigation was implemented, including comprehensive parametric analysis on both strategies mentioned above by implementing coupled neutronics and thermal-hydraulics calculations and transient analysis in accidental conditions.

This is mainly attributed to the reduction of core leakage along with the increase of heavy metal loading. On the other hand, one can obtain more superior performance of in situ utilization of thorium in the SEP schemes if the thorium residence time is lengthened sufficiently. Although this saving is smaller than the MOX case, the thorium loading requirement decreases dramatically with the increased thorium discharge burn-up, maximally one order of magnitude lower than the MOX schemes, indicating that the cost of thorium fuel can be minimized with relatively considerable uranium loading saving.

It is concluded that the SEP schemes are better choices for in situ utilization of thorium and minimizing the uranium loading requirement than the MOX schemes. Probabilistic safety assessment - Evaluation of dependency of human errors in PSA. Jan Prochaska. VUJE, a. Evaluation of human errors dependency plays important role in the PSA field. Importance of human factor including dependency follows from high portion of operator activities that are foreseen to be part of response on particular initiating events or to correct improper activity of safety systems including recovery actions.

More and more detailed PSA works imply the need to use an effective, robust and error prone way to cope with enormous cases leading to real or potential dependencies of human errors as well as to perform PSA quantification in such a way which avoids biasing PSA results as whole. The objective of this paper is to compare used methods to evaluate human errors dependency and analyze implementation how human errors dependency is implemented into PSA quantification process.

Paper also introduces approach used by VUJE to evaluate and quantify human error dependencies including self dependency as well as inter personal dependency.

Peter Schimann. Based on the performed national and EU-wide stress tests AREVA has implemented an additional procedure to define measure of increasing safety after the Fukushima accident. This additional initiated procedure so called robustness analysis. It is at least it should be a part of solving process related to the Fukushima-problem in order to perform safety studies before starting plant modifications.

AREVA has alredy performed a wide variety of safety studies, e. In the paper the effect of performed robustness analysis are be illustrated on the performed safety studies mentioned above. Reactor physics - Evaluation of the efectiveness of a simple GFR heterogenous control rod design. Physics and Technology, Ilkovicova 3, 19 Bratislava, Slovakia 2. The Generation IV International Forum GIF is a cooperative international endeavor that is trying to define and perform research and development needed to establish feasibility and performance capabilities of the next generation of nuclear energy systems.

This fast-spectrum reactor is a high temperature He cooled system operated in closed fuel cycle. The GFR design is featuring ceramic fuel and structural materials both allowing high temperatures and efficiency using helium coolant. One of the biggest concerns in terms of such a challenging system is the reactivity control system. The GFR reactor accommodates a system of control rods with highly enriched boron carbide. This system of 18 control and 13 diverse safety devices should provide appropriate reactivity worth for reactor regulation and shutdown.

The GFR reactor is still in development, therefore no final design of the control rods has been selected so far and only homogenous material compositions are available. Due to the shielding effects of the adjacent absorber pins it is likely that the worth of the heterogeneous control rod is lower than the worth of the homogenous one. This study deals with the proposal of a simple heterogeneous control rod design and with the evaluation of its effectiveness.

As the first step, the available homogenous design was studied using both stochastic and deterministic approaches to identify the absolute worth, the integral characteristics and the interferences of the control devices. On the basis of the achieved results a simple control rod design was proposed with an aim to achieve comparable reactivity worth with the homogenous design using exactly the same material composition and volume fractions of materials. In order to save absorber material the introduction of several neutron moderator materials as a part of control rod design was investigated.

A special part of this paper is dealing with the introduction of system of movable reflector which can serve as an additional safety system for accidental reactivity removal. In pressurized water reactors, the fuel loading pattern has significant meaning in terms of both safety and economics. An optimal loading pattern is defined as a pattern in which the local power peaking factor is lower than a predetermined value during one cycle and the effective multiplication factor is maximized to extract the maximum energy.

This analysis serves as an independent computational assessment of power density spatial distribution in the reactor core after the loading of new fuel type with higher average enrichment of 4. Due to the fact that new fuel type was implemented into the standard operation of NPP Mochovce, the local power peaking factors exceeded the design limits, especially for these assemblies. The origin of this phenomenon can be explained either by higher enrichment of uranium neutronics or by the insufficient flow of coolant thermo-hydraulic. The main objective of this analysis is to determine the power distribution across the core and the results comparison with the data obtained from the on-site power monitoring system in order to identify the principal source of power non-uniformity.

The isotopic compositions of the fuel assemblies in the investigated time steps were calculated by SCALE 6. The real detailed operational history of each subassembly in the one sixth of the core during the last three campaigns was used in the calculation to achieve the most reliable isotopic composition for the precise MCNP5 stochastic calculation. The incident neutron data libraries of temperature dependent cross sections were prepared by NJOY The analysis of the impact of power level uncertainty given by on site monitoring system to the calculated isotopic composition and the multiplication abilities of loaded assemblies were also carried out.

The paper gives a brief description of the geometrical and material models used in the calculations. According to the results, the calculated spatial distribution of the power density correlates with the power density distribution determined by on site power monitoring system. The calculation demonstrates the significant impact of the fuel assemblies with higher enrichment on the power density distribution in their vicinity, which was not sufficiently taken into account at the fuel loading pattern design. Westinghouse, Rue Montoyer 10, Bruxelles, Belgium 1. Oak Ridge National Laboratory, P.

The Consortium for Advanced Simulation of Light Water Reactors CASL was established in July for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. Determination and validation of fuel cooling with the use of isotopic factors. Those factors were obtained by applying the effect of fuel cooling and redefining the number densities for each burnup step and finally to correct the isotopic vector for each operational and cooling condition.

This was necessary to enhance the accuracy of the isotopic library in the CORD-2 package. Trkov, M. Computational study of neutron screens performance considering different absorbing materials. The upcoming development of the IV generation fast reactor systems requires the examination of the behaviour exhibited by the structural and fuel materials under the irradiation conditions prevailing in these reactors. The lack of operating fast reactors, hence the lack of capability to perform experiments in such irradiation environments can be compensated by creating the desired irradiation conditions in existing thermal research reactors, such as the Material Test Reactors MTRs.

This can be achieved via neutron shielding materials, the so-called neutron screens, which can be installed in specific irradiation locations of the above mentioned reactors. The effectiveness of various neutron screens in tailoring the neutron spectrum is here investigated, by examining different screen materials and thicknesses. The screens are here considered of cylindrical shape with a central hollow which hosts the irradiated target. The present work is a preliminary study intending to highlight the materials that could be effectively used as neutron spectrum tailoring media, with final target to specify the most appropriate neutron screen.

This will also require sensitivity studies on the screen thickness planned for the next stage of the work since the optimum screen configuration must combine a material with an absorbing capacity as high as possible with a screen thickness as thin as possible. Research reactors - Validation of the reactor model has been successfully shown in a separate study. Thus, as a subsequent step, the primary purpose of this work is to estimate the burnup value of the fuel rods by using Monte Carlo method and to compare with the recorded reactor data. Operating conditions e.

All the feedback effects are modeled accordingly. The analyses are carried out for the cases of fission product poison free and in equilibrium. Effective core multiplication factor, keff, is presented as a function of total energy generated as the reactor operates at a power level of kW and 10 W. The calculated core-averaged fuel burnup is found to be 0. In the light of current positions of control rods, this disagreement reveals that the current fuel loading have much more burnup than the recorded data suggest.

At last, the outcomes are discussed from the viewpoint of refueling strategy of the reactor core. This new core configuration needs to be properly characterized in order to support future research activities. Aim of this work is to present the results of the measurements of the in-core neutron flux distribution and energy spectrum performed applying a method based on the synergetic use of the Monte Carlo code MCNP and of a de-convolution technique of activated foils. The method allows to measure both slow and fast neutron components proving as result a neutron spectrum in energy groups.

The readings of ex-core nuclear instrumentation enable the safe operation of the reactor, give the operator the ability of precise reactor power monitoring and play a crucial role in the normalisation of reactor calculations. Application of best estimate plus uncertainty in review of research reactor safety analysis. Box LG80, Accra, Ghana. Energy literacy is an understanding of the nature and role of energy in the universe and in our lives. It is the ability to apply this understanding to answer questions and solve problems, especially regarding our energy future.

How does nuclear literacy fit into the broader energy literacy concept? The paper will bring insights into the communication process of developing the web project with details regarding: - the multidimensional content structure of the web-page, taking into account the importance of: o knowledge about energy, o links between energy and sustainability and o the necessary elements for discussing our energy future. This will be a first-of-the-kind presentation of the new web-page for NENE conference participants.

GEN energija d. Travelling Exhibition Fusion Expo. The Fusion Expo is a travelling exhibition where visitors could explore fusion energy. Fusion expo is presenting fusion energy as an environmentally acceptable, safe and sustainable energy source. Fusion research, technology and its future use are presented to the citizens of Europe.

Main target group of this exhibition are mediators such as journalists, teachers, decision makers, NGOs, students, taxpayers, voters, primary and secondary school pupils. Construction of new NPP is a large and extensive project which demands a lot of activities to be done in the preparatory phase of the project. Action plan provides an overview of important phases of the project and discuss short-term and mid-term activities arising therefrom.

The delineation of project schedule on 5 different phases and descriptions of each phase were prepared as well as description of project implementation schedule assumptions, major steps of project management like strategic decision-making process and establishment of a company for construction phase.

Action plan also establish project management system with associated procedures and software support. Key phases of action plan are planning and preparation phase, site evaluation and preparation, bidding process, construction and commissioning with preparation for operation phase. Rolando Calabrese. Complex decisions need besides economic utilization and technological practicability, the acceptability of the outcome to the various stakeholders. If the decision is not well accepted by society, it has little chance of successful implementation regardless of its economic and technical merits.

Most nuclear—related decision—making processes deal heavily with public opinion attitude. With this regard, different aspects may influence the public support for nuclear: demographic factors, knowledge, risks perception. Moreover, nuclear accidents have always changed public attitudes where a slow recovery was seen following the event.

The recent Fukushima accident has had a significant effect on the nuclear policies of many countries whose governments have changed or redirected their investments in nuclear energy. INPRO has established, for each area identified in the methodology, a set of requirements, organized in a hierarchy of basic principles, user requirements and criteria, including indicators and acceptance limits, that must be fulfilled to meet the overall target of sustainable energy supply.

Public information, participation, acceptance and political support issues are addressed in the infrastructure area under the statement of user requirement 3. Based on a review of the results published in the open literature, the approach of INPRO to the topics of public opinion and political support in the process of decision—making regarding the use of nuclear energy, is discussed.

Finally, moving from a hierarchical approach, feedbacks and inter—relationships in this area are proposed presenting preliminary results obtained by means of the Analytic Network Process. During inspection of fuel assemblies unloaded from reactor core several fuel assemblies with open defects were found as well as other fuel assemblies with tight defects.

The cause of open failure of fuel cladding was vibrations due to baffle jetting at the core baffle plate locations. Corrective measures were implemented prior to NPP core reload to prevent recurrence of such fuel defects in new fuel cycle. The regulatory body of Slovenia, the Slovenian Nuclear Safety Administration SNSA , regularly follows the conditions of fuel cladding integrity through a system of safety performance indicators.

Open fuel cladding defects were already diagnosed from reactor coolant activities since July , 14 months before the outage The SNSA also performed a major task of providing information on all activities and fuel conditions to the Slovenian public, NGOs and international expert organisations. The SNSA continues with monitoring of fuel conditions and assessment of corrective actions in new fuel cycle.

22nd Symposium of AER on VVER Reactor Physics and Reactor Safety

Sensitivity analyses to support determination of emergency planning zones around the nuclear power plant. In Slovenia the protective measures for population in case of a nuclear accident are determined by National protection and rescue plan for nuclear or radiological accidents that was upgraded last time in These bases comprise also the determination of the protective action zones around the NPP as well as predetermined measures for protection of population in these zones.

Recommended sizes of UPZ in modern references are larger than those defined in existing National plan 15 to 30 km. The paper will describe analyses that were performed at the Slovenian Nuclear Safety Administration to produce updated assessment of potential threat to the NPP vicinity in case of most severe accidents. Source terms used as input were selected based on IAEA recommendations. We used generic data of core isotopes inventory and design data on filters decontamination factors of the passive containment filtering venting system.

The filtered radioactive release is not reduced for the noble gases but it is times reduced for released Iodine isotopes and times for the released aerosols particulates. We used meteorological data of real weather conditions and simulated the dispersion of radioactive material with the RODOS code. Since this is a rather complex issue to be solved we will need to take into account much wider choice of input data for additional analyses that have to be performed.

By April a total of 41 reports and sub-reports were submitted. The SNSA reviewed all of the reports and provided comments. The review was focused on whether the PSR2 presented a complete review of a certain factor and whether eventually missing topics should be included. At these meetings the remaining open issues in the final revisions of the reports were discussed and by May most new revisions of the reports were submitted to the SNSA.

By December , all of the revised reports together with a summary report were delivered to the SNSA. The summary report contained a list of PSR2 recommendations with explanation of prioritization process for the action plan. PSR2 has proven that the plant safety is in accordance with modern standards and that the plant can operate safely for next 10 years. Implementation of the action plan in next 5 years represents the challenge to plant and SNSA. It will additionally improve the overall plant safety. Implementation of radiation protection relies on three principles: justification of practice, optimisation of exposures and dose limits.

While justification of practice and dose limits are somehow consequences of societal reasoning and decisions, optimisation of exposures is imposed as a requirement that should be satisfied by source users on the case-by-case and day-by-day basis. Of course, there is no explanation on how to reach it, but there must be a way since this is official requirement. Unfortunately, things are not as simple as it looks at first sight. First, ALARA is not about reaching the lowest possible doses, but about reasonably achievable low doses. It is a method of optimisation, where dose is not the only factor.

Author will also present some recommendations for application of ALARA principle in practices where source users do not have direct and continuous support of radiation protection experts. A process of bidding is one of the most important tasks that GEN energija will have to carry out before construction of the second unit of nuclear power plant JEK 2. It consists of several phases. The first phase of the bidding process is the preparation of bid invitation specifications BIS.

The following phases are the evaluation of the bids provided by the bidders and the contracting with the successful bidder. Based on IAEA the evaluation process generally taking not less than 6 months depending on available human resources and their qualification skills. The paper highlights a bid evaluation process with special attention to a bid technical evaluation process for JEK 2.

In addition to the technical including safety evaluation, the overall Bid evaluation process comprises economic, financial, contractual and other applicable aspects which have to be considered in the decision-making process of implementing the project and the selection of the supplier s.

This process starts with the receipt of the bids and ends with the issuance of the final evaluation report. A project organisation and process flow chart for the technical evaluation are presented and discussed. Analysis of Stratified Steam Explosions. When during a severe reactor accident the hot reactor core melt comes into contact with the coolant water a steam explosion may occur. During the steam explosion the energy of the molten corium is transferred to the coolant in a timescale smaller than the timescale for system pressure relief and so induces dynamic loading of surrounding structures.

A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and thus lead to the radioactive material release into the environment.


In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations. Stratified melt-coolant configurations, i. The main reason for this assumption of the low energetics of stratified steam explosions is based on the hypothesis that the amount of melt in the premixture formed in stratified configurations is insufficient to produce strong explosions. This hypothesis was based on analytical considerations that interfacial instabilities in a stratified configuration are not efficient in creating an explosive premixuture.

It was supported with data from experiments performed with mostly low temperature liquids, which showed rather low energy conversion efficiency and slow explosion propagation. However, the recently performed experiments in the PULiMS and SES KTH, Sweden facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts a considerable melt-coolant premixing layer formed.

The calculations were compared to available experimental data. Various simulation results will be presented and discussed. Ivo Kljenak. The Fukushima accidents also highlighted that both in-depth understanding of severe accident sequences and development or improvement of adequate severe accident management measures are essential in order to further increase the safety of NPPs operated in Europe.

The THAI experimental facility, located at Becker Technologies GmbH in Eschborn Germany , is basically a single-volume cylindrical vessel, with a volume of 60 m3, an internal height of 9. In the considered experiments, combustion with upward flame propagation occurred in air-hydrogen and air-steam-hydrogen atmospheres, at different temperatures, pressures and initial hydrogen concentrations. The calculated pressures, temperatures and flame propagation axial velocities are compared to experimental results and discussed.

Even today the Fukushima accident is challenging not only the operator and nuclear regulator in Japan but also the entire international and national nuclear safety framework. This framework is actually based on lessons learned from a relatively small number of nuclear accidents with wide consequences. As a result, major developments in various fields, e. One of the fields, which have been somehow put aside in the past, is the development of a national strategy applied in the post-accident phase. The development of the document was finalised in The Post-Accident Strategy is based on lessons learned from nuclear and other accidents, e.

The paper presents main components of the strategy in Slovenia. According to the document mentioned the Government could formally establish a team of experts, e. One of the main tasks of the team is to prepare so-called Rehabilitation Program reflecting actual situation. All areas must be incorporated in the Rehabilitation Program. Among them radiation protection requires specific attention as uncontrolled radiological situation is a unique characteristic of nuclear or radiological accidents and is strongly connected to all other areas.

The Post-Accident Strategy is in line with the ICRP recommendations incorporating flexibility when emergency and existing exposure situations are studied. For each strategic area the Post-Accident Strategy proposes a leading institution responsible for a specific measure and a list of national databases in order to control a situation on a long term. The document also identifies challenges ahead for institutions which might play a major role after a nuclear or radiological accident in Slovenia. The Post-Accident Strategy is the first attempt to a systematic approach to post-accident management in Slovenia.

The aim was to assess the capability of computer codes to model in an integral way the physical processes taking place during a severe accident in a pressurised water reactor, from the initial stages of core degradation, the fission product transport through the primary circuit and the behaviour of the released fission products in the containment. The FPT3 Benchmark was supported, with participation from 16 organisations in 11 countries, using 8 different codes.

The temperature history of the fuel bundle and the total hydrogen production, also taking into account of the H2 generated by the boron carbide oxidation were well captured, but no code was able to reproduce accurately the final bundle state, using as bulk fuel relocation temperature, the temperature of the first significant material relocation observed during the experiment. Total volatile fission product release was simulated, but its kinetics was generally overestimated, concerning the modelling of semi-volatile, low-volatile and structural material release, needs some improvement, notably for Mo, Ru for which it was observed a substantial difference between bundle and fuel release.

The retention in the circuit was not well predicted, this was due mainly to the boron blockage formation in the rising line of the steam generator, and the volatility of some elements Te, Cs, I could be better predicted. Containment thermal hydraulics are well calculated while as regards the containment aerosol depletion rate only the stand-alone cases provide acceptable results, whilst the integral cases tend to largely overestimate the total aerosol airborne mass.

Calculation of iodine chemistry in the containment turned out to be difficult. Its quality strongly depends on the correct prediction of chemistry speciation in the integral codes. The major difficulties are related to the presence of high fraction of iodine in gaseous form in the primary circuit, which affects the iodine behaviour in the containment.

In the benchmark a significant user effect was detected, i. This work reports the benchmark results comparing the main parameters, and summarises the results achieved and the implications for plant calculations. According to classifications adopted by the IAEA, small reactors are characterized by an equivalent electric output of less than MW while medium sized reactors by an equivalent electric power between and MW. Pressurized water small and medium sized reactors SMR generally adopt an integral layout of the primary circuit with in-vessel location of steam generators and control rod drives; one compact modular loop-type design features reduced length piping, an integral reactor cooling system accommodating all main and auxiliary systems within a leak tight pressure boundary, and leak restriction devices.

The SMRs as well as the advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts. Thus, to simulate correctly the main phenomena involved during an accident scenario, the coupling between primary circuit and containment has to be reproduced accurately, by using of different parameter to calculate the condensation rate of steam on the containment walls.

Furthermore, given that the containment plays a fundamental role during every accident scenario, it had to be taken into account just as a real safety system. The aim of this work is to validate the capability of the ASTEC code to simulate the entire accident event and the possible fission product releases in a SMR. Transition boiling modelling in the MC3D code. Fakulteta za matematiko in fiziko, Jadranska 19, Ljubljana, Slovenia 3.

The Generation IV International Forum has selected six technologies for future nuclear reactors, including innovative sodium cooled fast reactor. One of the important issues in core melt progression during a severe accident in the sodium cooled fast reactor is the likelihood and the consequences of a vapour explosion. A vapour explosion can occur when the hot core melt comes into contact with sodium. A strong enough vapour explosion in a nuclear power plant could jeopardize the reactor vessel and so lead to a direct release of radioactive material to the environment.

Several experimental programs were launched to help understanding and characterizing the vapour explosion phenomenon during the melt-sodium interaction. These experiments revealed an important effect of the sodium sub-cooling on the behaviour of the melt-sodium interaction. The vapour explosion probability and efficiency is decreasing with the sodium sub-cooling. The physical properties of sodium, which strongly affects the melt-sodium heat transfer, and the melt solidification, which strongly affects the energy efficiency during the explosion, are identified as the reason for the observed behaviour.

On the other hand, important analytical activities have been performed to study the vapour explosion phenomenon in light-water reactors. The analytical researches are performed with the use of comprehensive computer codes that are devoted to study the fuel-coolant interaction phenomenon. The models of the key processes that are implemented into the codes were developed manly with the focus on the water. Therefore the applicability of the codes and models for the fuel-sodium interaction must be demonstrated.

In this paper the transition boiling modelling with the MC3D code is analysed. For sodium, it is likely that boiling occurs essentially in the transition boiling regime and then the modelling of this regime is the most important. Indeed, film boiling regime should be reasonably reproduced with the models already existing for water. But the transition boiling regime is difficult to model, because this is essentially an unstable film boiling situation.

The purpose of the paper is to propose a model for the transition boiling regime. The first objective is to analyse experiments with water and sodium. In the experiments it was observed that the temperature range of the transition boiling regime strongly depends on the coolant sub-cooling. Additionally, in the experiments with the sub-cooled water, it was observed that a relative constant heat flux exists in the transition boiling region.

This is important to consider because the innovative sodium cooled fast reactor will operate at the sub-cooling of K. The second objective is to propose a model for a heat transfer and a temperature range for a relative constant heat flux. The model will be compared with the experimental data. Box , 10 Athens, Greece 1. Within the context of an operating nuclear reactor core, multi-physics calculations that account for the bonded nature of the neutronic and thermal-hydraulic phenomena are of major importance in reactor safety and design.

As regards the main key parameters concerning the coupled problems, the handling of the involved feedbacks between the two physical processes, the accuracy of the Monte-Carlo calculation and the convergence of such an iterative scheme, are the main ones. Another issue which should be considered carefully is the optimal, in terms of computational time, use of the neutronic Monte-Carlo code, since the main disadvantage of such codes is the high computational cost.

The results show that satisfying accuracy can be obtained in reasonable computational time when Monte-Carlo multi-processing is combined with proper selection of the Monte-Carlo parameters as well as the parameters of the coupled iterative scheme. Box 49, Budapest , Hungary 1. Future sustainable nuclear energy production worldwide with closed fuel cycle needs application of fast spectrum reactors cooled by liquid metal or gas. For this reason, several developments of the thermal hydraulic system and 3D reactor kinetic codes are going on.

Presently, their coupling became one of the most current tasks to be solved for performing safety analyses. Such types of coupled codes have been applied mainly for performing analyses of second and third generation reactors. Recently both codes have been further developed in the direction of the fourth generation research. It offers the possibility of choosing between different models for the simulation of fluid dynamics. The code structure is highly modular and allows an easy implementation of different physical models. Other independent modules e.

The GRS methodology is developed for pseudo 3D TH description of the core with mapping schemes fitted to the 3D neutronically simulated active core. A large number of benchmarks have been successfully solved with the coupled system code and many analyses have been performed for NPPs with different reactor types. In this paper the coupling of these upgraded codes and their verification are presented. The obtained k-eff value of the initial state is reasonably close to the benchmark value defined without feedback by supposing a constant average temperature of the fuel and the coolant.

The time dependent behavior of the obtained power, reactivity and temperatures showed that the thermal and the neutronic processes are influencing each other in the expected, reasonable way. Instability events have taken place in several BWR nuclear power plants, during normal operation or experimental test, due to the non-linear dynamic responses of these facilities. There are commonly three instability types observed in BWR: Control system instabilities, channel thermohydraulic instabilities and coupled neutronic-thermohydraulic instabilities.

The last type involves two feedback loops, on the one hand, the neutronic feedback, where reactivity changes are caused by variations in void fraction density , and on the other hand the thermohydraulic feedback, which affects the inlet flow rate. In this case the reactor core becomes unstable and starts to oscillate due to a positive pressure drop feedback because of density wave oscillations. Two modes of oscillations are observed within the core, the core-wide or in-phase oscillation and the regional or out-of-phase oscillation.

The out-of-phase oscillation, which is the topic of this work, the power of half core oscillates out-of-phase in relation to the other half. Since performing empirical test in nuclear power plants means great difficulty, as well as high costs, in order to simulate complex scenarios as those related to BWR instabilities, the use of coupled thermal-hydraulic and neutron kinetics system codes is a very useful option. In this nuclear power plant, four low-flow stability tests were realized in at the end of cycle 2 to characterize the unstable behavior of this NPP. The reactor core has been modeled with 72 thermalhydraulic channels, 71 representing the active core and 1 for the core bypass.

It has been chosen a mapping based in the lambda modes of the power with which the possibility of conditioning the oscillation pattern is avoided Moreover, it allows an acceptable accuracy without adding significant computational time. Romain Henry, Iztok Tiselj. Multi-physics and Multi-scale modelling is a task particularly challenging for the future. More precisely, the description of phenomena occurring in the core involves a coupling between neutronic and thermal hydraulics through the so called reactivity or thermal feedback. In water cooled reactor, thermal feedback and temperature coefficients of reactivity are important, when temperature change, microscopic cross section change impacting directly the neutron flux and so the thermal power which is an important parameter from an economic point of view.

Those reasons were naturally leading in the development of coupling method in order to predict correctly the behaviour of the core in normal operation and accidental situation. It exist two way to realise the coupling the first one is called internal coupling in which the code describing one discipline have a module able to solve equations from the other discipline. One classical example is the point kinetic model implemented in a thermal-hydraulics code. One should pay attention that the physical model of the module is a simple model obtain after approximations.

The coupling can also be extern, each discipline is described by its own code and a subroutine allows communications between them for data exchange. Prediction capabilities of our Coupled system are compared with the mentioned benchmark results. Overview of Fire PSA and supporting software.

  1. NEA - Publication.
  2. Operation and Performance Analysis of Steam Generators in Nuclear Power Plants.
  3. Preliminary program!
  4. Nuclear meltdown - Wikipedia.
  5. Peter Simurka. Continually increasing requirements on nowadays full scope PSA L1 and L2 as whole, which is multiplied by importance of specific data for all modes of operation of nuclear power plant, highlight role of input data used in PSA quantification process. This fact also emphasizes the role of capability to process all necessary information to analyze all nuclear plant modes by appropriate way.

    Even if above-mentioned aspects are relevant for all parts of nowadays PSAs, their importance is critical for internal hazards including specific fire analysis internal fire analysis constitutes one of the most challenging PSA tasks. Application of tailored information system forms one of the ways to speed up analyzing process, enhances manageability and maintainability of particular PSA projects and provides effective reporting mean to document process of work as well as traceable and human readable documentation for customers.

    This paper provides brief overview of VUJE approach and experience in this area. The paper introduces general purpose of database developed for fire PSA. Paper explains as this basic data source is enhanced by adding several relatively independent tiers to employ all common data for fire PSA purpose.