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Moscow, Russia

Tokmachev G.,JSC Atomenergoproekt
11th International Probabilistic Safety Assessment and Management Conference and the Annual European Safety and Reliability Conference 2012, PSAM11 ESREL 2012 | Year: 2012

The Atomenergoproekt engineering company, general designer of nuclear power plants (NPP), has been performing probabilistic safety assessments (PSA) for six NPPs in design. They belong to generations of advanced plants constructed or planned to be constructed in Russia, India, Turkey and Bulgaria including Generation 3+ plants. The main advantage of the NPP with a new generation reactor compared with Russian designs of previous generations is the use of advanced equipment and introduction of additional passive safety systems in a combination with conventional active systems. Implementation of diversity increases likelihood of the safety function fulfilment. The new VVER plants having inherent safety features are addressed in terms of their anti-Fukushima properties. The Fukushima accident has the impact on the PSA development itself. This is a challenge to PSA developers. Lessons learnt from the Fukushima event are to direct additional efforts to some points within the PSA development which are discussed. The paper is aimed at sharing some issues and experience gained from design modifications and PSA development for new advanced plants. Source

Kljenak I.,Jozef Stefan Institute | Kuznetsov M.,Karlsruhe Institute of Technology | Kostka P.,NUBIKI Nuclear Safety Research Institute | Kubisova L.,Nuclear Regulatory Authority of the Slovak Republic | And 3 more authors.
Nuclear Engineering and Design | Year: 2014

An experiment on hydrogen deflagration (Upward Flame Propagation Experiment - UFPE) was proposed by the Jozef Stefan Institute (Slovenia) and performed in the HYKA A2 facility at the Karlsruhe Institute of Technology (Germany). The experimental results were used to organize a benchmark exercise for lumped-parameter codes. Six organizations (JSI, AEP, LEI, NUBIKI, RSE and UJD SR) participated in the benchmark exercise, using altogether four different computer codes: ANGAR, ASTEC, COCOSYS and ECART. Both blind and open simulations were performed. In general, all the codes provided satisfactory results of the pressure increase, whereas the results of the temperature show a wider dispersal. Concerning the flame axial and radial velocities, the results may be considered satisfactory, given the inherent simplification of the lumped-parameter description compared to the local instantaneous description. © 2014 Elsevier B.V. All rights reserved. Source

Tokmachev G.,JSC Atomenergoproekt | Morozov V.,JSC Atomenergoproekt
Kerntechnik | Year: 2011

Customer requirements to probabilistic safety targets are usually stronger than existing Regulatory or IAEA ones. It appears that industry takes the lead over regulation in this case and forces the designer to find and implement appropriate means to enhance safety, which sometimes have no reference to practical experience. On the other hand, regulatory documents and the existing PSA methodology are mainly oriented to operating plants. This creates problems when developing a PSA as well as performing regulatory reviews. The scope of the PSA may be different depending on a design stage such as the development conceptual, basic or detailed design. In addition, the base case PSA is usually performed for NPP in design. However, a customer may require additional PSA applications to consider, for instance, risk monitoring. In this case the scope of the PSA should be extended to implement special attributes of the application needed that often requires specific information not available at the design stage. Lack of design information affecting PSA development may be associated with incompleteness of the design that is typical for interim design stages and communication problems caused by the involvement of many different companies in the deign activity. To deal with this issue bounding technologies and the iterative PSA development are used. However this sometimes contradicts to the "best estimate" approach recommended by regulatory guides. PSA development for advanced NPPs has raised some issues originated from unknown new components, processes and technologies incorporated into the design of an advanced plant. The paper addresses some issues resolved while carrying out PSAs for advanced NPPs. Some PSA results for new advanced VVER plants under construction and the first lessons learnt from the Fukushima accident are also discussed. © Carl Hanser Verlag, München. Source

Remizov O.V.,RAS Institute for Physics and Power Engineering | Morozov A.V.,RAS Institute for Physics and Power Engineering | Tsyganok A.A.,RAS Institute for Physics and Power Engineering | Kalyakin D.S.,RAS Institute for Physics and Power Engineering | And 3 more authors.
International Congress on Advances in Nuclear Power Plants 2010, ICAPP 2010 | Year: 2010

The design substantiation of the heat removal efficiency from Novovoronezh NPP-2 (NPP-2006 project with WER-1200 reactor) reactor core in case of primary circuit leaks and operation of passive safety systems only (among these are the systems of hydroaccumulators of the 1st and 2nd stages and passive heat removal system) has been performed on the base of computational simulation of the interrelated processes in the reactor and containment. Computational simulation has been performed in consideration of the negative non-condensable gases effect on steam generator (SG) condensation power. The main sources of non-condensable gases in the primary circuit are nitrogen arriving at the circuit, as hydroaccumulators of the 1st stage actuate and products of radiolysis of water. The feature of Novovoronezh NPP-2 passive safety systems operation is that during the system of hydroaccumulators of the 2nd stage (HA-2) emptying the spontaneous withdrawal of gas-steam mixture takes place from steam generators' "cold" headers into the volume of HA-2 tanks. The flow rate of gas-steam mixture during the operation of HA-2 system is equal to the volumetric water discharge from hydroaccumulators. The calculation carried out by the different integral thermal hydraulic codes have revealed sufficiency of such volumetric gas-steam mixture flow rate from the SG to HA-2 system so as to eliminate the " poisoning" of SG piping and to maintain necessary condensation power. To confirm the calculated investigation results, the experiment on a HA2M-SG test facility constructed at the IPPE has been carried out. The test facility incorporates WER reactor SG model with volumetric-power scale is 1:46. Steam to the HA2M-SG test facility is fed from the IPPE heat power plant. Gaseous addition to steam coming to the SG model is fulfilled from high pressure gas balloons. Nitrogen and helium, simulating hydrogen, are used in the experiments. The basic results of experimental studies aimed at determination of SG condensation power under the inflow to the tube bundle of gas-steam mixture and the simulation of gas withdrawal from steam generator's "cold" header to the HA-2 system are presented in the report. As a result of the research, carried out at the HA2M-SG test facility, the experimental substantiation has been obtained that in the case of emergency leak SG s have condensation power sufficient for effective heat removal by PHRS system from reactor. Source

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