Search results for: VVER

Authors: Pavel Borodkin, Nikolay Khrennikov, Azamat Gazetdinov

Abstract:

The problem of ensuring of VVER type reactor equipment integrity is now most actual in connection with justification of safety of the NPP Units and extension of their service life to 60 years and more. First of all, it concerns old units with VVER-440 and VVER-1000. The justification of the VVER equipment integrity depends on the reliability of estimation of the degree of the equipment damage. One of the mandatory requirements, providing the reliability of such estimation, and also evaluation of VVER equipment lifetime, is the monitoring of equipment radiation loading parameters. In this connection, there is a problem of justification of such normative parameters, used for an estimation of the pressure vessel metal embrittlement, as the fluence and fluence rate (FR) of fast neutrons above 0.5 MeV. From the point of view of regulatory practice, a comparison of displacement per atom (DPA) and fast neutron fluence (FNF) above 0.5 MeV has a practical concern. In accordance with the Russian regulatory rules, neutron fluence F(E > 0.5 MeV) is a radiation exposure parameter used in steel embrittlement prediction under neutron irradiation. However, the DPA parameter is a more physically legitimate quantity of neutron damage of Fe based materials. If DPA distribution in reactor structures is more conservative as neutron fluence, this case should attract the attention of the regulatory authority. The purpose of this work was to show what radiation load parameters (fluence, DPA) on all VVER equipment should be under control, and give the reasonable estimations of such parameters in the volume of all equipment. The second task is to give the conservative estimation of each parameter including its uncertainty. Results of recently received investigations allow to test the conservatism of calculational predictions, and, as it has been shown in the paper, combination of ex-vessel measured data with calculated ones allows to assess unpredicted uncertainties which are results of specific unique features of individual equipment for VVER reactor. Some results of calculational-experimental investigations are presented in this paper.

Keywords: equipment integrity, fluence, displacement per atom, nuclear power plant, neutron activation measurements, neutron transport calculations

Procedia PDF Downloads 133

Authors: Antoaneta Stefanova, Rositsa Gencheva, Pavlin Groudev

Abstract:

This paper presents an application of the ASTEC V2r3p3 computer code for simulation of QUENCH-12 experiment. The test has been performed to investigate the behavior of VVER type of fuel assemblies during severe accident conditions. In the performed analyses it has been assessed the mass of generated hydrogen during the experiment flooding of overheated core. The comparison of ASTECv2r3p3 calculated results with measured test data shows good agreement.

Keywords: hydrogen production, VVER, QUENCH facility, severe accident, reactor core

Procedia PDF Downloads 206

Authors: Antoaneta Stefanova, Rositsa Gencheva, Pavlin Groudev

Abstract:

This paper presents the modeling approach in SBO sequence for VVER 1000 reactors and describes the reactor core behavior at late in-vessel phase in case of late reflooding by HPIS and gives preliminary results for the ASTECv2 validation. The work is focused on investigation of plant behavior during total loss of power and the operator actions. The main goal of these analyses is to assess the phenomena arising during the Station blackout (SBO) followed by primary side high pressure injection system (HPIS) reflooding of already damaged reactor core at very late ‘in-vessel’ phase. The purpose of the analysis is to define how the later HPIS switching on can delay the time of vessel failure or possibly avoid vessel failure. For this purpose has been simulated an SBO scenario with injection of cold water by a high pressure pump (HPP) in cold leg at different stages of core degradation. The times for HPP injection were chosen based on previously performed investigations.

Keywords: VVER, operator action validation, reflooding of overheated reactor core, ASTEC computer code

Procedia PDF Downloads 392

Authors: Hygreeva Kiran Namburi, Anna Hojna, Edita Lecianova, Fencl Zdenek

Abstract:

Irradiation Assisted Stress Corrosion Cracking [IASCC] is one of the most significant environmental degradation in the internal components made from Austenitic stainless steel. This mechanism is still not fully understood and there are no suitable criteria for prediction of the damage during operation. In this work, core basket material 08Ch18N10T austenitic stainless steel acquired from decommissioned NPP Nord / Greifswald Unit 1, VVER 440-230 type, operated for 15 years and irradiated at 5.2 dpa is studied. This material was tensile tested at two different test temperatures and strain rates in air and at the elevated temperature under the water environment. SEM observations of the fracture surface documented ductile fracture of the samples tested in air, but areas of IASCC tested in water. This paper emphasizes on the microscopic examination results from the mechanically tested samples to determine the underlying IASCC physical damage process. TEM observations of thin foils made from the gauge sections that are closer to the fractured surface of the specimen aimed to find variances in interaction of dislocations and grain boundaries owing to different test conditions.

Keywords: irradiation assisted stress corrosion cracking, core basket material, SEM observations of the fracture surface, microscopic examination results

Procedia PDF Downloads 329