Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process

Analysis of dynamic processes during the accidents in a district heating system

Paper presented at the 9th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, Malta, 16-18 July, 2012.The accidents in the District Heating System are inevitable and they occur due to various reasons. Therefore it is necessary to perform the analysis of possible accident in piping system and to evaluate the consequences. After performing such analysis, it is possible to take the necessary measures to ensure safer and more reliable heat supply, so that the consequences of accidents are less severe. This paper demonstrated the capabilities of developed (using RELAP5 code) district heating network model for the analysis of dynamic processes. Three hypothetical accident scenarios in Kaunas city heating network are presented: (1) blackout in the Kaunas central part pump station; (2) break of heat supply pipe to northwestern district of Kaunas city; (3) rapid pump trip in one of Kaunas city pump stations. The discussion regarding dynamic processes (water hammer effect) in pipelines during the accidents is presented. The results of analysis demonstrated that the pressure pulsations as the accident consequences do not lead to the additional failures in pipelines in district heating system.dc201

Analysis of processes in spent fuel pools in case of loss of heat removal due to water leakage

Paper presented at the 8th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, Mauritius, 11-13 July, 2011.Safe storage of spent fuel assemblies in the facilities for intermediate storage (spent fuel pools) is very important. These facilities are not covered with a leak-tight containment, thus the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are very slow. Therefore, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to the leakage of water from a spent fuel pool. Also the accident mitigation measure, i.e. the injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal-hydraulic code RELAP5/MOD3 and the code for severe accident analysis ASTEC. The phenomena taking place during such accident are discussed.mp201

Application of COCOSYS code for investigation of gas mixing in mistra test facility

Paper presented at the 9th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, Malta, 16-18 July, 2012.In the case of a severe accident in a water-cooled nuclear power plant large amounts of hydrogen could be generated due to fuel claddings oxidation and released to the containment. At certain concentrations of steam air and hydrogen the hydrogen combustion could occur and challenge the structural integrity of the containment, which is a last barrier preventing from radioactive material release to the environment. Therefore, a detailed knowledge of containment thermal-hydraulics is necessary to predict the local distribution of hydrogen, steam and air inside the containment. This paper presents the experience of Lithuanian Energy Institute in simulation of the experiments performed in MISTRA test facility for the case of the International Standard Problem ISP47. The MISTRA facility is located in the Saclay center of France Atomic Energy Commissariat (CEA) and is related to the research of containment thermal-hydraulics and hydrogen safety. The MISTRA facility and its operating conditions are designed with reference to the containment conditions of a pressurized water reactor (PWR) in accident situation. The facility comprises containment inside which three condensers are set up and external circuits. Containment volume is ~100 m3, with an internal diameter of 4.25 m and a height of 7.3 m. Containment is not temperature regulated, but preheated by steam condensation and thermally insulated. The relevant physical phenomena for simulation are the following: 1)centered steam and helium (instead of hydrogen) injection in the containment; 2) pressure and temperature increase in the containment; 3) wall condensation at regulated wall temperature; and 4) flow pattern in the containment and resulting gas temperature and concentration distribution. Test sequence consisted of several transient and steady state stages, when the measurements of the gas temperature and gas concentration profiles where performed. The presented analyses were performed employing the code COCOSYS versions V2.0v2 and V2.3 developed at GRS mbH (Germany). COCOSYS is a lumped-parameter code for the comprehensive simulation of all relevant phenomena, processes and plant states during severe accidents in the containment of light water reactors. The free convection, forced convection, radiation heat transfer and condensation may be considered in the analysis. The condensation model is based on the heat and mass transfer analogy (Stefan’s law). The water and gas flows are calculated separately, i.e. different junctions have to be specified for these flows. Several zone models could be selected by the user. The EQUIL._MOD zone model assumes the perfect steam, gas and water mixture inside a zone. Each component of the mixture is in thermal equilibrium. NONEQUILIB model considers the water and gas mixture, which is not necessarily in thermal equilibrium, i.e. water and gas may have different temperatures and calculated separately in the energy balance. The experimental and analytical analyses showed that gas stratification was not observed and well-mixed atmosphere conditions were reached for the investigated case.dc201

Thermal characteristics of container for on-site irradiated nuclear fuel transportation

Paper presented at the 8th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, Mauritius, 11-13 July, 2011.An object of analysis in this paper is the container, which was developed for transportation of irradiated RBMK-1500 nuclear fuel assemblies at the Ignalina Nuclear Power Plant (NPP). Ignalina NPP (Lithuania) comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit 1 were safely transported and reused in the reactor of Unit 2 before final shutdown of the Unit 2 reactor in 2009. The RBMK reactor is continuously reloaded at power. Therefore the reactor core contains fuel assemblies with different burn-up level. After permanent reactor shutdown hundreds of fuel assemblies in the reactor core have considerably less burn-up than their design value. Such fuel assemblies have high energetic potential and can be reused. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 Fuel Assemblies for reuse in the Unit 2. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for NPP-s with RBMK type reactors. The structural integrity, thermal, radiological and nuclear criticality safety calculations were performed to assess the acceptance of the proposed set of equipment. The purpose of this paper is to present the results of thermal analysis of new developed container, which was used for transportation of irradiated RBMK-1500 nuclear fuel assemblies. Using finite element code the irradiated fuel transportation container model was developed and influence of an environment temperature and influence of different axial fuel power density profiles over container temperatures field was determined. Performed analysis demonstrated that the temperatures in proposed nuclear fuel transportation container do not exceed acceptance limits for both normal operation and accident conditions.mp201

Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process

Thermal–hydraulic performance of confinement system of RBMK in case of accidents

Within the framework of a European Commission sponsored activity, an overall assessment of the deterministic safety technology of the ‘post- Chernobyl modernized’ Reactor Bolshoy Moshchnosty Kipyashiy (RBMK) has been completed. The accident analysis, limited to the area of design basis accident, constituted the key subject for the study; therefore, the notorious Chernobyl Unit 4 event was outside the scope of the investigation. The present paper deals with the evaluation of the accident performance of the confinement of the RBMK. Ignalina-2 and Smolensk-3 Nuclear Power Plants (NPP) are considered. The documented activity includes four main parts: (a) description of key features of the confinement for the two NPP; (b) identification and characterization of relevant accident scenarios part of the (DBA) area; (c) key features and qualification level of adopted computation tools primarily including codes and input decks; (d) results from the analysis of selected accident scenarios. In the first part of the paper, the layout complexity of the RBMK confinement systems, including the building on the top of the reactor cavity is pointed out together with main differences between Smolensk-3 and Ignalina-2 units. Accident scenarios suitable for the analysis, second part of the paper, are derived from recently issued International Atomic Energy Agency (IAEA) reports, from the available Safety Analysis Reports (SAR) of existing RBMK NPP and from the list of bounding accidents derived from the already mentioned EC Project. The features of Cocosys and Relap5 thermal–hydraulic system codes developed in Germany and US, respectively, are outlined in the third part of the paper, giving emphasis to the qualification level in the range of parameters of interest to RBMK confinement systems. The adopted nodalisations resembling Ignalina 2 and Smolensk 3 NPP are also described including the demonstration of the qualification level. The results, fourth part of the paper, basically show the overall RBMK confinement system robustness within the DBA area. The important role of the reactor cavity is confirmed and of the venting. The number of broken channels causing the displacement of the upper reactor cavity plate is calculated in the range 20–25