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Groudev P.,Bulgarian Academy of Science | Petrova P.,Bulgarian Academy of Science | Kichev E.,Kozloduy NPP | Mancheva K.,Risk Engineering LTD
Risk, Reliability and Safety: Innovating Theory and Practice - Proceedings of the 26th European Safety and Reliability Conference, ESREL 2016 | Year: 2017

The article discusses options of using the Level 2 PSA results to support the application of Severe Accident Management Guideline (SAMG) in the Nuclear Power Plant (NPP). The efficiency of SAM depends on availability of appropriate equipment for accident management for all possible scenarios (NEA/CSNI 2010, IAEA 2012) and effective guides for the operators, which should be applied correctly. All these increase the importance of the training programmes for NPP operators to cope with severe accidents (NEA/CSNI 2002, IAEA 2006, IAEA 2009). The use of the Level 2 PSA in the NPP operator training, covering severe accident management can be considered in two main directions. © 2017 Taylor & Francis Group, London.

Georgieva E.L.,Risk Engineering Ltd. | Dinkov Y.D.,Risk Engineering Ltd. | Ivanov K.N.,Pennsylvania State University
Progress in Nuclear Energy | Year: 2014

A two-group cross-section generation methodology, cycle-specific cross-section update procedure and VVER-1000 reactor core model are described. The HELIOS lattice physics code is used to calculate the cross-section data tables according to a customized version of the cycle-specific cross-section modelling methodology of the Pennsylvania State University (PSU). A real-time version of the NEM code from PSU is developed for VVER-1000 full-scope simulator applications. The cross-section update procedure is tailored to meet cycle-specific reactor core simulation fidelity requirements as well as particular customer needs and practices. Combined with an enhanced thermal-hydraulics and Instrumentation (I&C) models the scope of Kozloduy 6 full-scope replica control room simulator is expanded to the whole range of plant operation modes ranging from cold shutdown (depressurized) state to rated power, as well as deviations from normal operating modes and even beyond-design basis accidents. The fidelity and accuracy of simulation is illustrated through comparison with plant-specific data. © 2014 Elsevier Ltd. All rights reserved.

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 | Year: 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.

Andonov A.,Risk Engineering Ltd. | Kostov M.,Risk Engineering Ltd. | Iliev A.,Bulgarian Academy of Science
Nuclear Engineering and Design | Year: 2015

The paper describes the procedure and the results from the assessment of the vulnerability of a generic pre-stressed containment structure subjected to a large commercial aircraft impact. Impacts of Boeing 737, Boeing 767 and Boeing 747 have been considered. The containment vulnerability is expressed by fragility curves based on the results of a number of nonlinear dynamic analyses. Three reference parameters have been considered as impact intensity measure in the fragility curve definition: peak impact force (PIF), peak impact pressure (PIP) and Momentum over Area (MoA). Conclusions on the most suitable reference parameter as well on the vulnerability of such containment vessels are drawn. The influence of the aircraft impact induced damages on the containment ultimate pressure capacity is also assessed and some preliminary conclusions on this are drawn. The paper also addresses a conceptual design of a protective structure able to decrease the containment vulnerability and provide a preliminary assessment of the applicability of such concept. © 2015 Elsevier B.V.

Kostov M.,Risk Engineering Ltd. | Henkel F.O.,Woelfel Beratende Ingenieure | Andonov A.,Risk Engineering Ltd.
Nuclear Engineering and Design | Year: 2014

The current paper presents key elements of the comprehensive analyses of the effects due to a large aircraft collision with the reactor building of Belene NPP in Bulgaria. The reactor building is a VVER A92; it belongs to the third+ generation and includes structural measures for protection against an aircraft impact as standard design. The A92 reactor building implements a double shell concept and is composed of thick RC external walls and an external shell which surrounds an internal pre-stressed containment and the internal walls of the auxiliary building. The malevolent large aircraft impact is considered as a beyond design base accident (Design Extended Conditions, DEC). The main issues under consideration are the structural integrity, the equipment safety due to the induced vibrations, and the fire safety of the entire installation. Many impact scenarios are analyzed varying both impact locations and loading intensity. A large number of non-linear dynamic analyses are used for assessment of the structural response and capacity, including different type of structural models, different finite element codes, and different material laws. The corresponding impact loadings are represented by load time functions calculated according to three different approaches, i.e. loading determined by Riera's method (Riera, 1968), load time function calculated by finite element analysis (Henkel and Klein, 2007), and coupled dynamic analysis with dynamic interaction between target and projectile. Based on the numerical results and engineering assessments the capacity of the A92 reactor building to resist a malevolent impact of a large aircraft is evaluated. Significant efforts are spent on safety assessment of equipment by using an evaluation procedure based on damage indicating parameters. As a result of these analyses several design modifications of structure elements are performed. There are changes of the layout of reinforcement, special arrangements and spatial reinforcement to increase shear resistance, as well as increase of the flexure reinforcement in most of the external protective structures. Significant requirements are formulated regarding the equipment stability; some requirements for the passive safety systems are adjusted and spatial arrangement of equipment are improved. All the investigated scenarios show that there are sufficient margins to prevent a severe accident. © 2013 Elsevier B.V.

Borisov E.,Risk Engineering Ltd. | Grigorov D.,Risk Engineering Ltd. | Mancheva K.,Risk Engineering Ltd.
Annals of Nuclear Energy | Year: 2013

This paper presents the results of analysis for application of a new Severe Accident Management Guideline (SAMG) approach which is specifically applied for VVER-1000/B320 reactor installations. In general, this innovative approach is fully applicable for all the pressurized water reactors from second generation. The purposes of the analysis for the new SAMG approach application are as follows: To represent suggestions for new engineering safety features application for SAMG strategies.To assess the applicability of the new engineering safety features and means for SAMG strategies in case of loss of all off-site power supply sources for VVER-1000/B320 reactor installations.To represent important operator actions and to analyse the effectiveness of these actions for accidents management in compliance with the new approach.The RELAP5/MOD3.3 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. The input data deck for the analysis is optimized, verified and validated. © 2013 Elsevier Ltd. All rights reserved.

Kostov M.,Risk Engineering Ltd.
Nuclear Engineering and Design | Year: 2014

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.

Svetlin P.,Risk Engineering Ltd
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2014

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.

Philipov S.,Risk Engineering Ltd | Filipov K.,Technical University of Sofia
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2014

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.

Philipov S.,Risk Engineering Ltd | Popov D.,Kozloduy NPP | Batakliev K.,HayCAD Infotech
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2014

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.

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