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.     

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

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.     

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.     

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.     

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.     

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.     

Comparative sensitivity study of the RBMK-1500 reactor one group distribution header blockage accident model

The article presents comparative sensitivity study of the Ignalina NPP RBMK-1500 reactor one group distribution header complete blockage accident model. The accident model was developed by RELAP5 thermal-hydraulic code and the accident scenario is one of the scenarios analyzed in the RBMK reactor safety analysis report. The sensitivity study comprised of the Fourier amplitude sensitivity test and the random sample based sensitivity measures (correlation coefficients and standardized regression coefficients). Two types of the model output were investigated: maximum temperature and dynamic temperature change during the progression of the accident. In addition, the effect of the input parameter distribution of different truncation levels and the effect of double standard deviation to the sensitivity results were studied.     

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.     

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.     

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.     

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 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.     

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.     

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.     

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.     

Analysis of the processes in spent fuel pools of Ignalina NPP in case of loss of heat removal

Ignalina NPP is the only nuclear power plant in Lithuania consisting of two units, commissioned in 1983 and 1987. Unit 1 of Ignalina NPP was shutdown for decommissioning at the end of 2004 and Unit 2 is to be operated until the end of 2009. Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500. According to the design, the spent fuel should be returned for reprocessing to Russia. However actually any fuel assembly has not been taken out from territory of the Ignalina NPP and all assemblies of spent fuel are stored in the spent fuel pools and dry on-site storage facility. Thus, the safety justification of facilities for intermediate spent fuel assemblies’ storage in Ignalina NPP is very important. This paper presents the results of loss of heat removal accidents (the most probable beyond design basis accident) in spent fuel pools of Ignalina NPP. The analysis was performed by employing best-estimate system thermal hydraulic code RELAP5 and codes for severe accidents ATHLET-CD and ASTEC. The best-estimate analysis, performed using RELAP5, allowed to investigate in the details the water evaporation, uncovering and fuel assemblies heat-up processes, when heat removal from the structures of buildings and pools are evaluated. The processes of spent fuel assemblies’ degradation due to loss of long-term heat removal were analyzed using ATHLET-CD and ASTEC codes. The results of calculations showed that the increase in water temperature in the pools from 50 °C up to 100 °C takes approximately 80-110 h, the evaporation of water volume down to uncovering of fuel assemblies takes approximately 220-260 additional hours. Later, after 200-300 h, the temperature of fuel claddings exceeds 800-1000 °C and the failures of fuel claddings occur due to cladding ballooning. The total amount of hydrogen generated up to time of complete water evaporation from spent fuel pools is about 7500-16,000 kg. These results of performed analysis were used for development of accident management guidelines for spent fuel pools of RBMK-1500.     

Development of accident management measures for RBMK-1500 in the case of loss of long-term core cooling

Results of the Level 1 Probabilistic Safety Assessment of the Ignalina Nuclear Power Plant have shown that in the risk topography transients are dominating. Analysis has shown that failure of the long-term core cooling is the main contributor to the core damage frequency. However, the reactor core damage in the long-term indicates the potential opportunities for the accident management. The main goal of accident management is to avoid a multiple fuel channel rupture because considering the design of RBMK reactors the consequences of rupture of more than 11–16 FC at full pressure would be close to the consequences of Chernobyl accident. This paper presents a detailed thermal–hydraulic analysis of the accidents with the loss of long-term core cooling, performed using the RELAP5 model of Ignalina NPP reactor cooling circuit and safety systems. Different ways of potential accident management are discussed. On the basis of this analysis the accident management strategy was developed.     

Reactor cavity and ALS thermal-hydraulic evaluation in the case of fuel channels ruptures at Ignalina NPP

This paper presents the work analysis of the thermal-hydraulic parameters behavior in the RBMK-1500 reactor cavity (RC) and other connected volumes in the case of fuel channels ruptures. The analysis is performed with CONTAIN code using the models of accident localization system (ALS) and reactor cavity venting system (RCVS). The RCVS capacity is assessed and expressed as a number of ruptured fuel channels at which the integrity of RC is maintained. The uncertainty analysis of pressure behavior in RC during multiple fuel channel rupture was performed. The initial and boundary conditions and the code models were selected and their influence on the results is estimated.Calculation of coolant mass and energy release to the reactor cavity in case of fuel channels rupture performed using the main circulation circuit model of Ignalina NPP, which was developed by employing state-of-the-art code RELAP5/MOD3.2 [Fletcher et al., RELAP5/MOD3 code manual user’s guidelines, Idaho National Engineering Lab., NUREG/CR-5535 (1992)]. These results were applied further as the initial data for the analysis of the thermal-hydraulic parameters behavior in the affected compartments employing CONTAIN code.     

Study of the nuclide inventory of operational radioactive waste for the RBMK-1500 reactor

The experimental and modeling results on the radionuclide concentrations in the Ignalina NPP operational waste are presented in the work. The scaling factors between easy-to-measure γ emitters ¹³⁷Cs, ⁶⁰Co and a number of difficult-to-measure radionuclides, the activity measurements of which are related to radiochemical procedures, α and β spectrometry, have been determined. The study shows that the scaling factor method can be applied for RBMK-1500 reactor waste characterization. The scaling factors were used in determination of the nuclide composition of operational radioactive waste and characterization of radioactive waste during the Ignalina NPP decommissioning.     

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.