Reliability of the Coolant Flow Rate Measuring System at the Ignalina Nuclear Power Plant

Age assessment and control of the reliability of the coolant flow-rate measurement system of the RBMK-1500 reactor at the Ignalina nuclear power plant are examined. A statistical analysis is made of the data obtained from measurements of the techanical characteristics of SHADR-32M flow meters. A mathematical model of the aging of flow meters and reliability estimates is constructed. A strategy is developed for replacing SHADR; this strategy makes it possible to support the required relaibility of the entire coolant flow-rate measurement system.     

Prediction of fuel channel—graphite gas-gap behaviour in RBMK reactors

To maintain thermal contact between the fuel assembly and the graphite moderator, RBMK design reactors employ graphite split rings, which are alternatively tight on the pressure tube or tight on the graphite brick central bore. The split in the graphite rings allows a helium/nitrogen gas mixture to flow up the fuel channel. This prevents oxidation of the graphite and can be sampled to detect pressure tube leaks. The initial clearance between the rings and pressure tube or graphite brick is approximately 2.7 mm (1.35 mm each side). Due to material property changes of the pressure tubes and graphite during operation of the reactor, the size of the clearance between the rings and the pressure tube/brick, called the “gas-gap”, varies. Closure of these gaps has been identified as a possible safety case issue by reactor designers and by independent reviews carried out as part of TACIS reviews and as part of the Ignalina Safety Analysis Report. The reasons for this are that gas-gap closure would cause the pressure tube to be tightly gripped by the graphite bricks via the split rings, which could lead to:

• Extra loading on the upper pressure tube zirconium/steel transition joint, particularly during shut down and emergency transients.

• Splitting of the graphite brick, leading to loss of thermal contact between the pressure tube and graphite. As approximately 5.6% of the heat in graphite-moderated reactor is generated within the moderator through neutron and gamma-heating, loss of thermal contact would result in higher graphite temperatures, accelerating the rate of graphite expansion and hence increasing the loading of the core radial restraint.

• Graphite debris may become lodged in inter-brick gaps, leading to increased axial pressure tube loading during shut down and emergency transients.

The authors have carried out deterministic assessments based on the Ignalina RBMK-1500 reactors in Lithuania, modelling the behaviour of the graphite under irradiation and have predicted graphite bore diameter changes that are in good agreement with the measurements of graphite bore diameters taken at Ignalina Nuclear Power Plant (NPP). A probabilistic model has been developed using the actual results of the deterministic calculations with non-linear graphite behaviour. Statistical analysis of the measurements of tube and graphite diameters taken from Units 1 and 2 at Ignalina NPP has been carried out. Further work has been carried out to try to determine the uncertainty inherent in the predictions of the gas-gap closure from the calculations. The overall objective of the studies is to aid prediction of the gas-gap closure process, and help to identify a suitable monitoring strategy for gas-gap closure that could be used for any RBMK reactor.     

Dynamic reliability and risk assessment of the accident localization system of the Ignalina NPP RBMK-1500 reactor

The paper presents reliability and risk analysis of the RBMK-1500 reactor accident localization system (ALS) (confinement), which prevents radioactive releases to the environment. Reliability of the system was estimated and compared by two methods: the conventional fault tree method and an innovative dynamic reliability model, based on stochastic differential equations. Frequency of radioactive release through ALS was also estimated. The results of the study indicate that conventional fault tree modeling techniques in this case apply high degree of conservatism in the system reliability estimates.One of the purposes of the ALS reliability study was to demonstrate advantages of the dynamic reliability analysis against the conventional fault/event tree methods. The Markovian framework to deal with dynamic aspects of system behavior is presented. Although not analyzed in detail, the framework is also capable of accounting for non-constant component failure rates. Computational methods are proposed to solve stochastic differential equations, including analytical solution, which is possible only for relatively small and simple systems. Other numerical methods, like Monte Carlo and numerical schemes of differential equations are analyzed and compared. The study is finalized with concluding remarks regarding both the studied system reliability and computational methods used.     

Improvement of the method used for hydraulic profiling of an RBMK-1500 core at the Ignalina nuclear power plant

An improvement of the method of hydraulic profiling of the RBMK-1500 core at the Ignalina nuclear power plant is examined. The hydraulic profiling method employs an approach where an adequate margin up to a crisis of heat transfer is provided in the channels of the reactor, the mass steam content at all channel exits is kept uniform, the number of adjustments of the stop-control valve at power is kept to a minimum, the flow rate is changed three times over the entire run of a fuel assembly, and the total flow rate through the reactor is maintained within the regulation limits over the entire run. The GIDRA software developed makes it possible to perform hydraulic profiling of RBMK-1500 using the method indicated. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 342–346, November, 2006.     

Estimation of the Gas Gap Between the Channels and the Graphite Masonry in an RBMK-1500 Reactor

One of the main factors limiting RBMK service life is the closure of the gas gap between the technological channels and the graphite masonry. The objective of the present work is to construct a mathematical model which estimates the probability of the presence of a gas gap during different periods of operation. To this end, a probabilistic model is constructed for predicting the diameter of the graphite masonry and the technological channels. Analysis showed that the model constructed for estimating the residual gas gap permits assessing with a high degree of confidence the process which changes the gas gap.     

Safety analysis of beyond design basis accidents in RBMK-1500 reactors

At present the design basis accidents for RBMK-1500 are rather thoroughly investigated. The performed analyses helped to develop and implement a number of safety modifications. Further plant safety enhancement requires developing emergency procedures that would enable beyond design basis accidents management by preventing core damage or mitigating consequences of severe accidents.     

Updating of risk-informed ISI programme for Ignalina RBMK-1500 nuclear power plant in Lithuania: Results and challenges

The paper presents a risk-informed in-service inspection (RI-ISI) pilot study project of 300 mm piping at Ignalina nuclear power plant (INPP) RBMK-1500 reactor, located in Lithuania. The RI-ISI study investigates optimal 300 mm piping ISI strategies with respect to risk and required resources. In total 1240 stainless steel welds were analyzed, assuming inter-granual stress corrosion cracking (IGSCC) to be the main damage mechanism. Pipe break frequency was estimated by probabilistic fracture mechanics methods and combined with safety barriers, provided by probabilistic safety assessment (PSA) study.After 3 years of operation, updating of RI-ISI was performed by taking into account new statistical data on pipe defects. Comparison with previous RI-ISI program was performed. The paper includes discussion on uncertainties in the study and robustness of RI-ISI programs.     

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.     

Advances of Barkhausen Emission Measurement

Barkhausen emission (BE) is a well-known phenomenon used in nondestructive testing techniques of ferromagnetic materials. The weaknesses of methods based on BE are related to the difficulties of ensuring measurement reproducibility because of the influence of the specimen on the measurement channel. In this paper, we propose a method of sensor calibration prior to measurement. To implement the calibration, a BE sensor is equipped with an additional calibration coil. The electrical pulse induced in the calibration coil is then used to reconstruct measured magnetic field pulses at the specimen surface.