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.     

Modeling of void, fast power and graphite temperature reactivity coefficients measurements for the validation of RELAP5-3D RBMK-1500 reactor model

This paper deals with the modeling of RBMK-1500 specific transients taking place at Ignalina NPP: measurements of void and fast power reactivity coefficients, as well as change of graphite cooling conditions transient. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and based on the obtained experimental results the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is unique and important from the point of view of model validation for the gap between fuel channel and the graphite bricks. The measurement results, obtained during this transient, enabled to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors.     

Revised calculation route for two group macroscopic cross-section library for RBMK-1500 type reactor

To increase the accuracy of predicted reactivity effects and coefficients for the unit equipped with a RBMK-1500 type reactor at Ignalina NPP, the calculation route used to generate the library of nuclear data constants applied in the neutronic/thermal hydraulic analysis has been updated with a modern version of the WIMS lattice code, WIMS8. The previously available two group library used with the QUABOX/CUBBOX-HYCA code, adapted to model the physical and nuclear processes in a RBMK-1500 reactor core, was created using the freely available WIMSD reactor physics cell code and its associated nuclear data library. In this article, the results that are obtained under the performance of the new two group cross-section library generated with WIMS8 for RBMK-1500 design core are presented. This discussion is mostly concentrated on the prediction of the key physics parameter for the RBMK type reactor, the void reactivity coefficient, as this parameter has been underestimated, especially at higher fuel irradiation.     

Validation of coupled neutronic/thermal-hydraulic code RELAP5-3D for RBMK-1500 reactor analysis application

This paper summarizes RELAP5-3D code validation activities carried out at the Lithuanian Energy Institute, which was performed through the modeling of RBMK-1500 specific transients taking place at Ignalina NPP. A best estimate RELAP5-3D model of the INPP RBMK-1500 reactor has been developed and validated against real plant data, as well as with the calculation results obtained using the Russian STEPAN/KOBRA code. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters, as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data, which demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors. Future activities are discussed.     

Analysis of decay heat removal from RBMK-1500 reactor in decommissioning phase by natural circulation of water and air

The main problem in nuclear energy is providing of safety at all stages of lifetime of nuclear installations in conditions of normal operation, accidents and at shutdown. Ignalina NPP, located in Lithuania, is one of the latest with RBMK reactors at highest capacity. Ignalina NPP has two units, both are closed for decommissioning now (in 2004 and 2009). Both units are equipped with RBMK-1500 reactors, the thermal power output is 4200 MW, the electrical power capacity is 1500 MW for each. In RBMK-1500 reactor the fuel assemblies remain for long time inside reactor core after the final shutdown. The paper discusses possibility of heat removal from the RBMK-1500 core at shutdown condition by natural circulation of water (1) and air (2) inside the fuel channels. In first case the decay heat from fuel assemblies is removed due to natural circulation of water and the piping above reactor core should be cooled by means of ventilation in the drum separator compartments. To warrant free access of air in to fuel channels (in the second case) the reactor cooling system should be completely dry out and the pressure headers and the steam discharge valves in steam lines should be opened. If mentioned conditions will be fulfilled, the reactor core will be cooled by natural circulation of water or air and fuel rods remain intact.     

Determination of I activity in the RBMK-1500 main circulation circuit

The purpose of this work was to evaluate the content of difficult to measure isotope ¹²⁹I in the RBMK-1500 reactor fuel-to-clad gap and reactor main circulation circuit (MCC) coolant. To determine fission product (FP) release from the defective fuel, the methodology proposed by Lewis and Husain for the CANDU reactor primary coolant activity prediction was applied. The determined effective diffusion coefficient D′ = 1.2E−09 s⁻¹ of iodine in the RBMK-1500 fuel is higher than the one evaluated for the CANDU fuel 6.8E−10 s⁻¹. Results show that the method developed by Lewis and Husain can be applied for the RBMK-1500 fuel gap and reactor main circulation circuit coolant activity prediction.     

Analysis of accidents leading to local flow stagnation in parallel fuel channels of RBMK-type reactors

The RBMK-type nuclear power reactors, still operating in Russia, are graphite-moderated with vertical fuel channels, using low-enriched nuclear fuel. The main challenge, which leads to the overheating of the fuel assemblies, fuel channels and other core components in channel type nuclear reactors, is a misbalance between heat generation in core structures and heat sink, which can appear due to the loss of coolant accident. In this accidental case, the emergency core cooling system ensures the core cooling. In RBMK-type reactors this system consists of hydro-accumulators and a number of pumps, taking water from large water reservoirs. This equipment injects water into the fuel channels through the group distribution headers at high pressure. However, the direct supply of cold water from emergency core cooling system into fuel channels is possible only if check valves on group distribution headers are closed properly. If these check valves failed, the part of water would be lost through the break, the flow stagnation in channels could occur, which might lead to overheating of fuel assemblies in the fuel channels. This paper presents the results of deterministic safety analysis, performed using system thermal hydraulic code RELAP5. Using this code the reactor cooling system of RBMK-1500 was modelled and analyses of loss of coolant accidents with failure of few check valves in group distribution headers were performed. The results of the calculations are used for the development of symptom-based emergency operating procedures for RBMK-1500. The basic principles that describe the complex distribution of water flows in vertical forced circulation circuit with parallel fuel channels can be adjusted for the RBMK-1000 reactors, still operating in Russia.     

Modelling of the spatial distribution of the induced activities in the RBMK-1500 reactor graphite blocks and rings/sleeves

There is one nuclear power plant (NPP) in Lithuania – the Ignalina NPP – which is under decommissioning now. The Ignalina NPP has two units with RBMK-1500 reactors, which are the most powerful and the most advanced versions of RBMK-type reactor design. Unit 1 of the Ignalina NPP was shut down at the end of 2004 and Unit 2 was shut down at the end of 2009. RBMK is a water-cooled graphite-moderated channel-type power reactor and the decommissioning of these reactors faces specific challenges for proper characterisation and disposal of irradiated reactor graphite.Apart from radiological inventory, the spatial distribution of radionuclides in the reactor graphite is also very important because it could indicate the possibilities for decontamination/treatment of the irradiated graphite. This is important for consideration of the near surface disposal option for irradiated graphite, as without treatment it usually does not meet the waste acceptance criteria.Based on that, the work presented in this paper is focused on the modelling of the induced activity spatial distribution in the Ignalina NPP RBMK-1500 reactor graphite components: blocks and rings/sleeves. The modelling was performed with MCNP and SCALE computer codes and consisted of two mains stages: modelling of the neutron flux in the reactor graphite components, and then modelling of the neutron activation in them using the already modelled neutron flux. In such a way, the spatial induced activity distribution in the analysed reactor components was obtained. Modelling results show that the thermal neutron flux is more intensive in the outer radial regions of the graphite components and this, in general, results in higher induced activities there.     

Specifics of RBMK core cooling in beyond design basis accidents

The most dangerous beyond design basis accidents for RBMK reactors, leading to the worst consequences, are related to the loss of long-term heat removal from the core. Due to a specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using control rods cooling circuit, depressurisation of reactor cooling system, supply of water into cooling system from low pressure water sources, etc. This paper presents the analysis of such heat removal by employing RELAP5, RELAP5-3D and RELAP/SCDAPSIM codes. The analysis was performed for Ignalina nuclear power plant with RBMK-1500 reactor. The analysis of result shows that the restoration of water supply into control rod channels enables to remove 10-30 MW of the generated heat from the reactor core. This amount of removed heat is comparable with reactor decay heat in long-term period and allows to slowdown the core heat-up process. However, the injection of water to reactor cooling system is considered as main strategy, which should be considered in RBMK-1500 accident management procedure.     

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.     

Benchmarking analysis of water hammer effects using RELAP5 code and development of RBMK-1500 reactor main circulation circuit model

Correct prediction of water hammer transients is of paramount importance for the safe operation of the plant. Therefore, verification of computer codes capability to simulate water hammer type transients is a very important issue at performing safety analyses for nuclear power plants. Verification of RELAP5/MOD3.3 code capability to simulate water hammer type transients employing the experimental investigations is presented. Experience gained from benchmarking analyses has been used at development of the detail RELAP5 code RBMK-1500 model for simulation of water hammer effects in reactor main circulation circuit. In RBMK-type reactors the water hammers can occur in cases of rapid check valve operation. The performed analysis using RELAP5 code RBMK-1500 model has shown that in general the maximum values of the pressure pulses due to water hammer do not exceed the permissible loads on the pipelines.     

Method based on isotope ratio mass spectrometry for evaluation of carbon activation in the reactor graphite

The general idea of this work is to introduce an evaluation method to restore the irradiation parameters of graphite or other carbonaceous materials using experimental and modelling results of ¹³C generation in the irradiated material. The method is based on coupling of stable isotope ratio mass spectrometry and computer modelling of the reactor core to evaluate the realistic characteristics of the reactor core such as the neutron fluence in any position of the reactor graphite stack or other graphite constructions.The generation of carbon isotopes ¹³C and ¹⁴C in the irradiated graphite of the RBMK-1500 reactor has been estimated by modelling of the reactor core with computer codes MCNPX and CINDER90. Good agreement of simulated and measured Δ¹³C/¹²C values in graphite of the central part of the reactor core indicates that the neutron flux (1.40 × 10¹⁴ n/cm² s) is modelled accurately in the graphite sleeve of the fuel channel. The simulated activity of ¹⁴C is compared with the one measured by the β spectrometry technique. Results indicate that production of ¹⁴C from ¹⁴N in the RBMK-1500 reactor is considerable and has to be taken into account in order to make proper evaluation of ¹⁴C activity. Measured ¹⁴C specific activity values correspond to 15 ± 4 ppm impurity of ¹⁴N in graphite samples from the RBMK-1500 reactor core.     

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.     

Analysis of BDBA in RBMK-1500 reactor with long-term loss of heat removal from the core

The Ignalina nuclear power plant (NPP) is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. The accident management guidelines for beyond design basis accidents (BDBAs) are in a stage of preparation at Ignalina NPP. The most challenging event from BDBAs is the unavailability of water sources for heat removal from fuel channels (FCs). Due to specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: depressurisation of reactor cooling system (RCS) (if pressure in cooling circuit is high) and supply of water into cooling system from low pressure water sources, removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using cooling circuit of control and protection system channels, etc. The possibility to remove the heat using cooling circuit of control and protection system channels looks very attractive, because the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. The heat from fuel channels, where heat is generated, through graphite columns is transferred in radial direction to cooled channels with control rods. Therefore, the heat removal from RBMK-1500 reactor core using control rods cooling circuit can be used as non-regular mean for reactor cool-down in case of BDBAs with loss of long-term heat removal from the core.     

Was the Chazhma Accident a Chernobyl of the Far East?

The 1985 reactivity accident on a submarine in bukhta Chazhma was accompanied by a substantial emission of fission products and activational radionuclides whose total activity reached 5 MCi. Some specialists have compared this emission to the emission resulting from the 1986 accident in Chernobyl, neglecting the large difference in the radionuclide composition: short-lived products of prompt fission of uranium (with an admixture of activational ⁶⁰Co) in Chazhma and long-lived fission products accumulated over the run of the power reactor with an admixture of short-lived nuclides from the spontaneous excursion of the RBMK-1000 reactor in the Chernobyl nuclear power plant. It is shown that the emission of long-lived radioecologically significant radionuclides in Chazhma was approximately 0.79 Ci, while in the Chernobyl accident this emission was 90 MCi.A quantitative comparison is presented of the activity and radionuclide composition between the accidents in Chazhma and Chernobyl taking account of the fraction of long-lived radionuclides and neglecting the radioactive inert gases. These quantitative estimates are used to show that the Chazhma accident is not analogous to the 1986 accident in Chernobyl.     

Development and adoption of measures for increasing the safety of power-generating units of nuclear power plants with RBMK reactors in 1986–2005

The basic stages and directions of upgrading and reconstruction of the power-generating units of nuclear power plants with RBMK reactors over a period of 20 yr since the Chernobyl accident are reflected in this paper. The greatest amount of work was done in 1986–1990 on all units with RBMK-1000 and-1500 reactors which were operating at that time in the USSR. The purpose of the upgrade was to improve the neutron-physical characteristics of the reactor cores, increase the response speed of the safety and control system and decrease the positive effect of water outflow from the cooling loop of this system, increase the flow capacity of the systems performing emergency discharge of the steam-gas mixture from the reactor, and improve the basic operating documentation. The subsequent stages of the upgrading and reconstruction concerned mainly the first-generation power-generating units (Nos. 1 and 2 units of the Leningrad and Kursk nuclear power plants). These works were performed to make the power-generating units conform to the requirements of the modern normative documents on safety and allow for the possibility of these units to remain in operation after the nominal 30-yr service life has been exhausted. __________ Translated from Atomnaya énergiya, Vol. 100, No. 4, pp. 312–320, April, 2006.     

Analysis of relative release rates of radionuclides from the RBMK-1500 reactor fuel elements

Specific activities (concentrations) of fission products (FP) and activation products in spent fuel elements of the RBMK-1500 reactor were calculated using SCALE 5 computer code. Different burnup (5.1–21.0 MWd/kg) fuel assemblies were experimentally investigated. Activities of radionuclides present in the coolant water of storage cases of defective fuel elements were experimentally measured and analyzed. Experimental results provide a basis for a quantitative analysis of radionuclide release from spent fuel of the RBMK-1500 reactor. Relative release rates of radionuclides from the fuel matrix were assessed based on a comparison of experimental results with theoretical calculations. On the basis of analysis results released fission and activation products can be divided into several groups according to their release rates from fuel; this can be generalized for radionuclides with similar chemical properties.     

Uncertainty and sensitivity analysis of MCPs’ trip events at Ignalina NPP

Eight main circulation pumps (MCPs) are employed for the cooling water forced circulation through the RBMK-1500 reactor at the Ignalina nuclear power plant (NPP). There have been a few events when one or more MCPs were inadvertently tripped.This paper presents investigation of a one MCP trip event and all MCPs’ trip events at Ignalina NPP. Thermal-hydraulic analysis was conducted using the best estimate system code RELAP5/MOD3.3. Uncertainty and sensitivity analysis of flow energy loss in different parts of the main circulation circuit (MCC), initial conditions and code-selected models was performed. Such analysis allows to estimate the influence of separate parameters on the calculation results and find those modelling parameters that have the largest impact on the investigated events. Uncertainty analysis indicates that natural circulation provides adequate cooling in the case of all MCPs tripped, and that the reactor is reliably cooled by forced circulation in the case of a single tripped MCP. On the basis of this analysis, recommendations for the further improvement of model are developed.     

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.     

A model based on bootstrapped neural networks for computing the maximum fuel cladding temperature in an Rmbk-1500 nuclear reactor accident

The present paper illustrates the use of artificial neural networks for computing the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in an RBMK-1500 nuclear reactor. The uncertainties associated to the neural predictions are quantified by resorting to the bootstrap technique. The trained neural networks are further used to perform a sensitivity analysis aimed at identifying the parameters which most significantly influence the maximum fuel cladding temperature.