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Sofia, Bulgaria

Groudev P.,Bulgarian Academy of Science | Petrova P.,Bulgarian Academy of Science | Kichev E.,Kozloduy NPP | Mancheva K.,Risk Engineering Ltd.
Safety and Reliability of Complex Engineered Systems - Proceedings of the 25th European Safety and Reliability Conference, ESREL 2015

The Nuclear Power Plant (NPP) safety shall be ensured through consistently applying the defence in depth concept based on the use of a system of physical barriers to the release pathways of ionising radiation and radioactive substances to the environment, as well as on a system of technical and organizational measures (levels) to protect the barriers and retain their effectiveness and to protect the population, the personnel and the environment. NPP safety shall be analysed using deterministic and probabilistic methods to verify and confirm the established design basis and the effectiveness of defence in depth arrangements. Is there any interconnection between deterministic and probabilistic safety assessments to support the decision making to insure the NPP safety in case of severe accident? Is there the Probabilistic Safety Assessment (PSA) any contribution to development and application of Severe Accident Management Guidelines (SAMG)? The authors of this article are trying to pose again these questions in the light of the event in the Fukushima Dai-ichi NPP and the results of the stress tests. It is an attempt to present the whole range of NPP modes, including normal operation, Anticipated Operational Occurrences (AOOs), Design Basis Accident (DBA), Design Extended Conditions (DECs), and severe accidents with their frequencies and documents covered (e.g. FSAR, QM, EOPs, SAMG). The possible contribution of PSA to the main task of development of SAMG, based on IAEA safety guide NS-G-2.15 and WENRA reference levels are proposed. © 2015 Taylor & Francis Group, London. Source

Kostov M.,Risk Engineering Ltd.
Nuclear Engineering and Design

There are continuous attempts to describe the damaging potential of the seismic (or vibratory) motion by a single parameter or a set of damage indicating parameters (DIP). Recently CAV and IJMAJMI have become two very promising parameters. Originally CAV was introduced as a parameter alongside spectral characteristics of ground motion for assessment of the operational state of a nuclear power plant after a seismic event. The aim is to shorten the time for evaluation of OBE exceedance and to provide guidance for the quick restart of a seismically affected plant. Meanwhile, there has been a growing experience and confidence that the DIP could be used not only as global indicator, i.e. assessment of the severity of excitation on the plant site but also as damage descriptor at equipment level, i.e. at each equipment location. The procedure proposed is similar to that for floor response spectra generation and safety evaluation against seismically induced forces. The current paper presents basic relations between damage parameters and structural damage derived from the European strong motion database. The seismic experience database is utilized to assess the capacity/damage of equipment. A formalized approach is considered for evaluation of critical facilities subjected to dynamic vibratory loading. The following sequence of evaluation steps is discussed: Step one: for the safety equipment the standard in-structure CAV is calculated and compared with a threshold to screen-out the equipment for further considerations. An additional and optional threshold could be the in-structure IJMA intensity estimate. Step two: for all locations where standard CAV of in-structure vibrations is higher than the threshold, the floor response spectra are evaluated. They have to be compared with the equipment capacity spectra. The latter are represented by design floor response spectra multiplied by a safety factor or seismic ruggedness spectra. Step three: alternatively or simultaneously with the ultimate capacity assessment (force driven design) a displacement based evaluation of the ultimate drift capacity of the respective equipment can be performed. It has to be stressed that under high frequency excitation the displacement (drift) estimated capacity is by far more realistic than the force based estimates. If none of the above checks is positively answered detailed conventional analysis can follow; however, a much smaller amount of equipment would remain for assessment. © 2013 Elsevier B.V. Source

Philipov S.,Risk Engineering Ltd. | Popov D.,Kozloduy NPP | Batakliev K.,HayCAD Infotech
International Conference on Nuclear Engineering, Proceedings, ICONE

This paper presents the results of comparison between CFD code FloEFD and an experiment in Kozloduy NPP concerning coolant mixing. This comparison is done with purpose to find out behavior of FloEFD to manage complex tasks. Real experimental data is used from a performed experiment in the frame of an international project on nuclear safety. The main assessment point is the temperature distribution of the coolant at core inlet. A representative parameter is considered axis of minimal mixing. Also, temperature distribution in the down chamber at level 2.5m from the bottom is assessed. Copyright © 2014 by ASME. Source

Philipov S.,Risk Engineering Ltd. | Filipov K.,Technical University of Sofia
International Conference on Nuclear Engineering, Proceedings, ICONE

This paper presents the results of an analysis of the application of CFD tool to help hydrogen management. Some information pointed out the problem of hydrogen generation and distribution. Passive autocatalytic recombiners are the point of interest and mainly PAR units' location. A severe accident is taken into account regarding the sources of hydrogen generation. The analysis of the severe accident progression is performed with MELCOR code. CFD tool Fluent (ANSYS) is applied to assess hydrogen and steam distribution in the atmosphere of the containment (confinement). The NPP unit of type WWER 440 (V230) is considered but as it is stressed this fact is irrelevant to phenomena and accident management targets. Copyright © 2014 by ASME. Source

Svetlin P.,Risk Engineering Ltd.
International Conference on Nuclear Engineering, Proceedings, ICONE

Initiating events such as primary to secondary loss of coolant (PRISE) can lead to conditions forming reversed flow from the second to the primary circuit. Current issue shows the results of a CFD analysis of the distribution of boric acid on the entrance of the core in case of such reversed flow of coolant as a result of PRISE initiation event. Analyzed accident is included in the list of design basis accidents and requires precise approach in analyzing the phenomena associated with the possibility of injection of coolant with low concentration of boric acid in the primary side. The paper emphasizes on the application of CFD to solve the problem. Analyzing the accident is done in advance with the help of system code RELAP. The input data as flow rate, concentration and temperature at the inlet of the reactor is submitted as boundary conditions in FLUENT and boric acid mixing is analyzed to the core inlet. Copyright © 2014 by ASME. Source

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