Egidijus Urbonavičius

Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

The Specifics of RBMK Core Cooling at Overheated Conditions

Reactor RBMK-1500 of Ignalina NPP is a boiling light water reactor with graphite moderator. Several important design features of RBMK-1500 are unique and extremely complex in respect to western reactors: the fuel assemblies are loaded into individual channels rather than a single pressure vessel; the plant can be refueled on-line; the neutron spectrum is thermalized by a massive graphite moderator. The reactor coolant system consists of two loops, each having flow length of more than 200 m. There are 1661 of vertical parallel fuel channels and numerous components, such as headers, pumps, valves, etc. The fuel channels and fuel claddings are made of Zirconium-Niobium alloy. From the point of view of safety barriers, each fuel channel in RBMK-type reactor corresponds to a pressure vessel of vessel-type reactors. Thus, the fuel channels are the most important element in reactor cooling system. However, in case of beyond design basis accidents with loss heat removal from the core the integrity of fuel channels could be challenged as they are not so strong as the pressure vessel. The paper presents the analysis of different possibilities to cooldown the core of RBMK-type reactors. Injection of water to RCS is considered as main strategy. Such “bleed and feed” procedure is used for vessel type reactors, but at present is not considered at RBMK-1500.© 2006 ASME

RALOC4 code validation against measured data at Ignalina NPP during single Main Safety Valve opening

Abstract Although RALOC4 code is validated against many experiments with regard to Western Nuclear Power Plants (NPPs) the code validation problem for the Accident Localization System (ALS) of Ignalina NPP modeling is of special importance because the condensing pools at NPP with RBMK-1500 differ from the pressure suppression systems installed in NPPs with German BWR. The response of Ignalina NPP ALS to the unintentional opening of single Main Safety Valve, which occurred in 1998, is analyzed by employing code RALOC4. The results of post-event calculations compared with the measured data available after the event. The performed analysis showed that RALOC4 code could be applied for the simulation of Ignalina NPP ALS. Nevertheless, the spray modeling in RALOC4 should be improved allowing the simulation of sprays in NON_EQUILIBRIUM zone model and to consider the diameter of water droplet diameter and height of droplet fall.

Accession of Lithuanian energy institute to nuclear fusion researches

Nowadays there is extremely appreciable increasing energy demand. The main public problem is to find new ways to cover energy needs, and also do not forget influence of climate change and declining supplies of fossil fuels by tackling this problem. Future energy stations must be clean, carbon-free and be able to generate large amounts of qualitative energy. Nuclear fusion energy stations have a number of advantages: unlimited fuel, no CO2 and no air pollution, an avoidance of major accidents, no radioactive “ash” and no long-lived radioactive waste, competitive electricity generation cost and other. Nuclear fusion energy stations have many advantages, but they also have some disadvantages as well. First of all, there is a little explored range and, therefore, there is a need for more accurate researches and new development in nuclear fusion area. The gas of hydrogen - deuterium and tritium is heated to very high temperature (~108 K) and at this temperature plasma must be maintained as long as possible in nuclear fusion reaction. These physical and technical problems are very complicated, and in order to achieve efficient use of energy from nuclear fusion reaction these issues should be solved. Scientists from Lithuanian energy institute (LEI) also participate in the international fusion projects. Lithuanian energy institute is one of the members in the European fusion project (Eurofusion) from 2004. LEI research areas are most of all related to the investigation of Wendelstein 7-X experimental fusion, stellarator type, and device. LEI scientists perform safety assessment of the fusion facilities in the field of thermal-hydraulics, neutron transport analysis and system analysis.

Project design progress on plasma and outer vessel exhaust gas system based on LOCA safety analysis of W7-X stellarator

Wendelstein 7-X (W7-X) is a fully optimized low-shear stellarator and shall demonstrate the reactor potential of this fusion plant. It is presently under construction at the Greifswald branch institute of IPP. The cryostat of W7-X stellarator has a torus shape which consists of a plasma vessel as an inner vessel, the outer vessel and the ports, which connect the plasma vessel (PV) with the outer vessel (OV). In operating mode several postulated hazard events which could occur on W7-X stellarator were identified. A set of actions were initiated, in order to mitigate or minimize the possible damaging consequences especially for both main W7-X vessels (OV and PV). The document presents the requirements of the exhaust gas safety systems currently under design for both vessels with overpressure protections working in failure mode.

Post-benchmark simulation of the THAI+ facility TH27 experiment

ABSTRACT In order to maintain the integrity of a nuclear power plant containment and effectively manage a severe accident, it is necessary to understand phenomena occurring in the atmosphere of the nuclear power plant containment during the accident. A number of containment atmosphere mixing experiments have been performed in the dedicated experimental facilities, followed by the numerical simulations using lumped-parameter and computational fluid dynamics codes. This paper presents the THAI+ test facility experiment TH27 post-benchmark simulations performed with the lumped-parameter code ASTEC. The experiment TH27 was an initial operation test of the THAI+ facility, which has been recently constructed by expanding the experimental facility THAI with the newly constructed parallel attachable drum vessel. The experiment featured steam and helium injections and transport and mixing of gasses and steam between the two vessels, as well as wall heating and cooling of different vessels. The TH27 experiment was performed together with an international multistage benchmark, consisting of double-blind, blind, and open phases. The developed nodalization scheme and the features of the calculation are presented in the paper. The results of the calculations are compared to the experimental values for the main containment parameters – pressure, gas and wall temperatures, helium concentrations.

Simulation of Hydrogen Distribution in Ignalina NPP ALS Compartments During BDBA

Accident Localisation System (ALS) of Ignalina NPP is a “pressure suppression” type confinement, which protects the population, employees and environment from the radiation hazards. According to the Safety Analysis Report for Ignalina NPP ∼110 m3 of hydrogen is released to ALS compartments during the Maximum Design Basis Accident. However in case of beyond design basis accident, when the oxidation of zirconium starts, the amount of generated hydrogen could be significantly higher. If the volume concentration of hydrogen in the compartment reaches 4%, there is a possibility for a combustible mixture to appear. To prevent the possible hydrogen accumulation in the ALS of the Ignalina NPP during an accident the H2 control system is installed. The results of the performed analysis derived the places of the possible H2 accumulation in the ALS compartments during the transient processes and assessed the mixture combustibility in these places for a beyond design basis accident scenario. Such analysis of H2 distribution in the ALS of Ignalina NPP in case of BDBA was not performed before.Copyright

Simulation of ignalina NPP accident localisation system behaviour employing the RALOC4 code in the case of group distribution header rupture

The response of the RBMK Accident Localisation System (ALS) to a medium break LOCA (group distribution header (GDH) rupture) is analysed employing the RALOC4 code. The analysis employs a best estimate mass/energy source and considers both short and long-term (up to 24 h) responses of the ALS. The main long-term energy sources and sinks are taken into account and modelled in detail in this analysis. The technical systems connected to the ALS are taken into account and modelled in detail in this analysis in order to achieve correct mass and energy balances. The analysis results demonstrate that the pressures in the ALS compartments are below the maximum allowed design pressures in the case of GDH rupture.