Trnava, Slovakia
Trnava, Slovakia

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Bujan A.,VUJE Inc. | Ammirabile L.,Institute for Energy and Transport of the Netherlands | Slaby J.,VUJE Inc.
Progress in Nuclear Energy | Year: 2011

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic safety studies as a major contributor to core melt frequency and source term risk. The fission products release and transport to the containment has been analyzed during mid-loop operation of a reference PWR 1000 MWe reactor using the severe accident integral code ASTEC V2.0. The analyses have been performed considering the loss of residual heat removal (RHR) system at various times after reactor shutdown for the reactor vessel configuration with the removed upper head (open reactor). In this configuration, the possible air ingress can have an impact on safety such as accelerated oxidation and increased volatility of certain FPs (particularly iodine and ruthenium). Sensitivity calculations have been performed in terms of air ingress simulation with a different intensity. Besides equilibrium chemistry model, most of the calculations have also used a limited kinetics model. The study has shown that without air ingress the only predicted gaseous form of iodine is HI (≤7.4% of the total mass of iodine released from core) and no gaseous RuO4 is created. Sensitivity calculations have illustrated that the gross fraction of gaseous iodine (I2 + HOI + HI) has an increased trend with growth of air ingress intensity and with the duration of sequence evolution. In most oxidative atmosphere the gross iodine gaseous fraction could increase by a factor form of two to several times as compared to the corresponding case without air ingress (particularly due to I2 persistence). Creation of gaseous RuO4 is sensitive to carrier gas temperature; therefore a considerable fraction (≤3%) is predicted only in the sensitivity cases with the shortest time of loss of RHR after reactor scram. © 2011 Elsevier Ltd. All rights reserved.

The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident. This paper presents preliminary results of the numerical simulation of the WWER440/V213 reactor vessel internals (RVI) dynamic response to maximum hypothetical Large-break Loss of Coolant Accident (LOCA). The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such permanent (plastic) deformations occur in the RVI which would prevent timely and proper activation of the emergency control assemblies. In the case of the LOCA accident it is assumed rapid "guillotine" break of one of the main coolant pipes and rapid depressurization of the primary circuit. The pressure wave spreads at the speed of sound, enters the reactor pressure vessel and causes deformation and stress in reactor vessel internals. The finite element model was created by MSC.Patran (Patran, 2010) and dynamic response was solved using MSC.Dytran (Dytran, 2008) finite element code. The model consists of reactor vessel internals (Lagrangian solid elements) and water coolant (Euler elements) inside the reactor. Arbitrary Lagrangian Eulerian (Belytschko et al.; 2003) coupling was used for simulation of the fluid-structure interaction. The calculation assumes no phase change in the water. No comparison with the experiment was performed up to now, because the required experimental data are not accessible for this type of the reactor. The most important acceptance criteria for the reactor internals demands that the movement of the emergency control assemblies under all operating conditions including accident is ensured (BNS, 2008). The numerical simulation of the WWER440/V213 reactor internals response to a LOCA accident showed that the acceptance criteria for RVI is fulfilled and required NPP safety standards are satisfied. © 2010 Elsevier B.V.

Breza J.,VUJE Inc. | Breza J.,Slovak University of Technology in Bratislava | Darilek P.,VUJE Inc. | Necas V.,Slovak University of Technology in Bratislava
Annals of Nuclear Energy | Year: 2010

The first step in investigation of thorium fuel is evaluation of the results obtained from the spectral code for this type of fuel. The benchmark summarized by IAEA in 2003 was used for partial validation of the code HELIOS 1.9. The benchmark was focused on a comparison of the methods and basic nuclear data. Acceptable results of benchmark comparison allowed examining and comparing different advanced nuclear fuel cycles under light water reactor conditions, especially in VVER-440. Cycles, calculations and results for VVER-440 reactors are presented in the paper. Two of the investigated thorium based fuels include one solely plutonium-thorium based fuel, while the other one is a plutonium-thorium based fuel with a content of reprocessed uranium. The third examined fuel cycle is a cycle with an inert-matrix fuel consisting of reprocessed plutonium and minor actinides (MA) fixed in an yttria-stabilized zirconium matrix. All of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The Pu transmutation rate and cumulating of Pu with MA in the spent fuel were compared mutually and with an UOX open cycle. The fuel cycle with an inert-matrix fuel was proven to be the best cycle for minimizing the production of Pu in the VVER-440 reactors. © 2010 Elsevier Ltd. All rights reserved.

Figedy S.,VUJE Inc.
7th International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies 2010, NPIC and HMIT 2010 | Year: 2010

In today's NPPs there is a demand for development of improved core surveillance methods with the ability to detect and diagnose possible unanticipated operational problems. The purpose of this work is to present the work status in development of new validation approaches of in-core sensors, both self powered neutron detectors and thermocouples, mainly based on implementation of computational intelligence techniques. The promising candidate is the OECD Halden Reactor Project's neuro-fuzzy system PEANO, which has originally been developed for various process parameters validation, like power, pressures, temperatures, flows, water levels etc. To use it for incore sensors signal validation is a challenging task, as its training would normally have to be done prior to each fuel cycle for ever changing fuel load patterns and bum-up of both fuel and self powered neutron detectors emitters. To overcome this difficulty, the idea arose to train PEANO to residuals, i.e. the differences between core simulator physics code results and experimental measured values. Thus, in residuals the reactor core physics is implicitly contained. The other approach based on combination of correlation coefficients and mutual information indices, which reflect the correlation of signals in linear and nonlinear regions, is mentioned only briefly. The preliminary results show that there is a potential to make the in-core sensor validation more reliable by integrating various methods based on computational intelligence and advanced signal processing innovative techniques together with the classical approaches.

Figedy S.,VUJE Inc.
World Scientific Proc. Series on Computer Engineering and Information Science 7; Uncertainty Modeling in Knowledge Engineering and Decision Making - Proceedings of the 10th International FLINS Conf. | Year: 2012

In this work an alternative method of in-core sensor validation is outlined. Instead of using direct sensor readings, it is based on neuro-fuzzy modeling of residuals between the experimental values and their theoretical counterparts obtained from reactor core simulator calculations. Throughout the fuel cycle, the neural networks are subject to the incremental training on data prior to the current monitoring period. While for thermocouples the method shows very good sensitivity, for self powered neutron detectors the results are not satisfactory. The purpose of this paper is to present the newest results from considerably long time effort in development of this computational intelligence based approach.

Kovacik J.,VUJE Inc.
Proceedings of the 8th International Scientific Symposium on Electrical Power Engineering, ELEKTROENERGETIKA 2015 | Year: 2015

Power transformers are important parts of substations from financial point of view. Unplanned outage of transformer represent of very high cost in addition. Presentation is dealing about a history of installation online monitoring system of power transformers in industry of Slovakia as well as experiences from their service.

Chrapciak V.,VUJE Inc.
ICNC 2015 - International Conference on Nuclear Criticality Safety | Year: 2015

The fuel suppliers improve permanently fuel for commercial energetic nuclear reactors. One way is to increase initial enrichment of U235. Higher initial enrichment of U235 allows to increase burnup of assembly, to prolong stay in assembly in reactor and therefore decrease number of fresh assemblies loaded into reactor and decrease number of spent fuel assemblies for storage and disposal. In this article are some criticality analyses for fresh fuel PWR 17x17 and VVER-440 for enrichment up to 6% of U235 without burnable absorber: - Criticality of fresh assembly (PWR 17x17 and VVER-440) in cold state in reactor - Critical number of fresh assemblies (PWR 17x17 and VVER-440) in water - Possible storage and transport of fresh fuel VVER-440 in existing facilities and transport cask The SCALE 6.1.2 system was used for all calculation. The KENO VI module was used with library v7-238 (continuous energy).

The complexity of the emergency preparedness and post-accident management was recognised and analysed in depth as well as relations among different stakeholders, and their roles and tasks within the post-accident preparedness process. Each stakeholder defined its specific role and work, and forms a part of the complex system; each stakeholder defined where and how to be involved in the whole system. Recovery was defined in the Slovak legislation by one general word before the process of the post-accident preparedness started. The process revealed the recovery issues and started the discussion in depth. Communication between different stakeholders involved in the active work of the Slovak stakeholder panel at facilitated workshops, seminars, workshops, training courses and exercises was found to be very important to get a balanced view on various aspects of the issues at stake at the national, regional and local levels. It enabled a common language and a shared understanding of the challenges to be developed. © EDP Sciences 2016.

Michal V.,VUJE Inc.
3rd Int. Joint Topical Meeting on Emergency Preparedness and Response and Robotics and Remote Systems 2011, EPRRSD, and 13th Robotics and Remote Systems for Hazardous Environments | Year: 2011

The paper deals with remotely operated and robotics technologies for physical and radiological characterization, retrieval, dismantling, demolition, decontamination and other activities during implementation of the most significant decommissioning projects of European nuclear power plants and other nuclear facilities. Overview of technologies that are developed and used mainly in Belgium, France, Germany, United Kingdom, Slovakia etc. is done. Due to the limited available information only the basic one concerning relevant technologies developed and used in Ukraine (for NPP Chernobyl) and in Russian Federation is provided. Brief history of utilization of remotely operated and robotics technologies for the implementation of relevant decommissioning activities in Europe is described including basic technical background of relevant technologies, the current status and expected future trends for their development and deployment. European approach is compared with the current state-of-the-art worldwide applications. Lessons learned and future trends are summarized in conclusions.

Kubacka J.,VUJE Inc.
IYCE 2013 - 4th International Youth Conference on Energy | Year: 2013

The objective of this paper is to present the preliminary results of design basis reconstitution of the systems, dedicated to prevention of containment boundary failure of the VVER440/V213 unit. The results are based on large set of calculations (using MELCOR code) of various Loss of Coolant Accident scenarios, which have been integrally evaluated and processed to preliminary set of parameters (characteristics) of bubble tower and containment spray system. An accent will be given to the comparison of existing characteristics of relevant systems to the by design basis reconstitution derived ones. The paper will provide recommendations which may lead to the modifications of some parameters of the safety systems. © 2013 IEEE.

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