Eugenijus Uspuras

Benchmark analysis of main circulation pump trip events at the Ignalina NPP using RELAP5 code

The state-of-the-art code RELAP5/MOD3 was originally designed for PWRs. Because of unique RBMK designs the application of this code to RBMK-1500 encountered several problems. A successful best estimate RELAP5 model of the Ignalina NPP has been developed. This model includes the reactor main circulation circuit (MCC) and reactor control and protection system required for this kind of transient analysis. Benchmark analysis of all operating main circulation pump (MCP) trip events was performed. During the analysis the characteristics of isolation control valves and MCP throttling regulating valves were established. Comparison of calculated and measured parameters was also used to establish realistic resistances of different MCC components and realistic behaviour of the controllers of the reactor systems. Calculations performed with the RELAP5 model, which includes these modifications, compare favourably with plant data.

State-of-the-art computer code RELAP5 validation with RBMK-related separate phenomena data

Abstract Validation of the RBMK model, developed by employing best estimate system computer code RELAP5 is performed by employing the data from NPPs operation or from integral and separate effects facilities. Validation of the models on the basis of separate phenomena is necessary to perform due to the fact that RELAP5 code has been developed for PWRs, which operate at different conditions (pressure, temperature, coolant void fraction, etc.) from RBMKs. In addition to that, there is a number of phenomena specific for RBMK type reactors (oscillatory flow rate behaviour in parallel channels, flow stagnation in channels, stratification in long horizontal piping, etc.), which have not been studied during RELAP5 validation for PWRs. In the paper, RELAP5 models for separate effects related to RBMK-1500 are presented and modelling of transients is performed. Obtained results are compared with experimental data.

Thermal-hydraulic analysis of accidents leading to local coolant flow decrease in the main circulation circuit of RBMK-1500

There are a few transient and loss-of-coolant accident conditions in RBMK-1500 reactors that lead to a local flow decrease in fuel channels. Because the coolant flow decreases in fuel channels (FC) leads to overheating of fuel claddings and pressure tube walls, mitigation measures are necessary. The accident analysis enabled the suggestion of the new early reactor scram actuation and emergency core cooling system (ECCS) initiation signal, which ensures the safe shutdown of the reactor and compensates the stagnation flow. Analysis of such conditions is presented in this paper. Thermal-hydraulic analysis was conducted using the state-of-the-art RELAP5 code. Results of the analysis demonstrated that, after implementation of the developed management strategy for destruction of local flow stagnation, the Ignalina nuclear power plant (NPP) would be adequately protected following accidents, leading to local coolant flow decrease in the primary circuit.

Thermal–hydraulic evaluation of the RBMK-1500 accident confinement system using CONTAIN 11AF

The response of the RBMK Accident Confinement System to a large break LOCA, medium break LOCA and small break LOCA is analyzed using the CONTAIN 11AF code. The effect of Condenser Tray Cooling System failure is investigated for the large break LOCA case. The analysis employs a best estimate mass/energy source and considers both short and long-term responses of the Accident Confinement System. Parametric studies are performed to evaluate the effects of water deposition on the short-term pressure peak and of by-pass leakage on long-term pressure increases.

Development of Ignalina NPP RBMK-1500 reactor RELAP5-3D model

Abstract This paper deals with the development of an integrated thermal-hydraulics–neutronics model for RBMK-1500 reactors for the analysis of specific plant transients in which the neutronic response of the core is important. A successful best estimate coupled RELAP5-3D model of Ignalina nuclear power plant (NPP) has been developed. The validation of the thermal-hydraulic model has been performed using operational transients from Ignalina NPP. The results of the calculations obtained with the RELAP5-3D model compare reasonably with the real plant data. The RELAP5-3D nodal kinetics model provides reasonable agreement with Ignalina NPP reactor power and coolant density profiles. The eigenvalue is close to unity, indicating that reasonable values are calculated for the neutron fluxes.

Analysis of potential waterhammer at the Ignalina NPP using thermal–hydraulic and structural analysis codes

The RBMK (Russian acronym for ‘channeled large power reactor’)-1500 reactors at the Ignalina nuclear power plant (NPP) have a series of check valves in the main circulation circuit (MCC) that serve the coolant distribution in the fuel channels. In the case of a hypothetical guillotine break of pipelines upstream of the group distribution headers (GDH), the check valves and adjusted piping integrity is a key issue for the reactor safety during the rapid closure of check valve. An analysis of the waterhammer effect (i.e. the pressure pulse generated by the valves slamming closed) is needed. The thermal‐hydraulic and structural analysis of waterhammer effects following the guillotine break of pipelines at the Ignalina NPP with RBMK-1500 reactors was conducted by employing the RELAP5 and PipePlus codes. Results of the analysis demonstrated that the maximum values of the pressure pulses generated by the check valve closure following the hypothetical accidents remain far below the value of pressure of the hydraulic tests, which are performed at the NPP and the risk of failure of the check valves or associated pipelines is low. Sensitivity analysis of pressure pulse dependencies on calculation time step and check valve closure time was performed. Results of RELAP5 calculations are benchmarked against waterhammer transient data obtained by employing structural mechanics code BOS fluids.

Development and evaluation of additional shutdown system at the Ignalina NPP by employing RELAP5 code

Abstract The anticipated transients without scram (ATWS) study in the ‘in-depth safety assessment of Ignalina NPP’ (SAR) showed that some ATWS scenarios can lead to unacceptable consequences. The apparent lack of effective inherent safety features for RBMK reactors leads to one high priority recommendation: that a second fast acting, independent and diverse reactor shutdown system needs to be installed. Another compensatory measure—the additional shutdown system (DAZ)—which has the potential to reduce the overall risk has been recommended for implementation at Ignalina NPP until the second shutdown system is installed. An analysis that demonstrated the suitability of the DAZ system—selected input process parameters and setpoints values—has been performed. The thermal-hydraulic analysis of loss of preferred electrical power and loss of both turbines with failure of the existing control and protection system (CPS), but with activation of the DAZ system, showed that the reactor is adequately protected by the DAZ system in these cases and acceptance criteria are not violated.

RELAP5-3D code validation in the neutron-dynamic analysis of transient processes taking place in RBMK-1500 reactors

The paper presents an evaluation of RELAP5-3D code suitability to model-specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian complex neutronic-thermal-hydraulic code STEPAN/KOBRA, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN/KOBRA codes, showed reasonable mutual agreement of the calculation results of both codes and their reasonable agreement with the real plant data.

Ignalina NPP RBMK-1500 gas gap evaluation

This paper presents Ignalina NPP Unit 1 RBMK-1500 reactor core lifetime analysis. The closure of the gas gap between the pressure tubes and the graphite bricks is one of the criteria for the evaluation of the reactor core lifetime. The rate of closure of the approximately 1.5 mm gaps between the pressure tubes and the graphite is largely a function of accumulated fast neutron dose and graphite operating temperatures. The main task of this paper is development of strategy and methodology for gas gap closure evaluation.

Integrated failure probability estimation based on structural integrity analysis and failure data: Natural gas pipeline case

In this paper, the authors present an approach as an overall framework for the estimation of the failure probability of pipelines based on: the results of the deterministic-probabilistic structural integrity analysis (taking into account loads, material properties, geometry, boundary conditions, crack size, and defected zone thickness), the corrosion rate, the number of defects and failure data (involved into the model via application of Bayesian method). The proposed approach is applied to estimate the failure probability of a selected part of the Lithuanian natural gas transmission network. The presented approach for the estimation of integrated failure probability is a combination of several different analyses allowing us to obtain: the critical cracks length and depth, the failure probability of the defected zone thickness, dependency of the failure probability on the age of the natural gas transmission pipeline. A models uncertainty analysis and uncertainty propagation analysis are performed, as well.