Ignalina NPP Unit Research Articles

Characterisation of RBMK-1500 graphite: A method to identify the neutron activation and surface contamination terms

In this study, we provide the results of radiological characterization of the RBMK-1500 graphite using modeling and nuclear spectrometry; we also introduce a method to identify neutron activation and surface contamination terms for the spent graphite waste. The simplified model of the reactor core created by programme package SCALE 6.2 for 4 × 4 fragment of the core was used for simulation of neutron activation of graphite impurities in the RBMK-1500 reactor. Calculations were supplemented by the non-destructive measurements of gamma-ray emitting nuclides and destructive analysis of selected samples of the graphite stack of Ignalina NPP Unit 1 reactor. Our analysis demonstrates that the partial contribution of different contamination sources can be identified by combining modeling and measurements. The findings on radionuclides such as 14C, 36Cl, 60Co, 134Cs, 137Cs, 154Eu as well as actinides in Ignalina NPP Unit 1 graphite having a significant impact on graphite activity and radiological characterization are discussed. The proposed method is also applicable for identification of contamination source in the other activated components of the reactor core.

Assessment of the Radioactive Contamination of the RBMK-1500 Reactor’s Steam Pipelines and High Pressure Rings

Data analyses of radioactive contamination of the RBMK-1500 reactor’s steam pipelines (SP) and components of high pressure rings (HPR) are presented in this paper. Also, modelled results of the SP-HPR system are compared to the results of other RBMK-1500 systems at Ignalina NPP Unit 1. Characteristics of SP-HPR components, thermal-hydraulic conditions of the coolant, and system operational regimes were evaluated employing the computer code LLWAA-DECOM (Tractebel Energy Engineering, Belgium). The presented results complement radiological characterization activities and facilitate the decommissioning process of nuclear facilities with RBMK type reactors. Analysis of the modelled results showed that the spread of radioactive contamination is very uneven between different components of the SP-HPR. The overall activity level of deposits of the SP-HPR is mostly determined by activated corrosion products and is lower than the activity level in the main circulation circuit (MCC) and nonpurified water subsystem activity of the purification and cooling system (PCS).

Uncertainty of determination of 158Tb in the RBMK nuclear reactor waste

The activity of 158Tb was measured in waste samples from the Ignalina NPP Unit I RBMK-1500 reactor using gamma-ray spectrometry. The origin of 158Tb and the other observed gamma-ray emitters has been studied by using SCALE 6.1 modeling and comparing radionuclide ratios in the RBMK-1500 radioactive waste. The results of the calculation of the massic activity of gamma-ray emitters were used for interpretation of the total gamma-ray spectrum and the determination of 158Tb massic activity uncertainty in the waste of RBMK-1500.

Analysis of the radiological contamination of the RBMK-1500 reactor low salted water system

The paper presents modelling and analysis results of contamination of the RBMK-1500 reactor Low Salted Water System (LSWS) components and comparison of the results with other RBMK-1500 water systems at Ignalina NPP Unit 1. These results provide information for radiological characterization, which is an important task of the nuclear installation to be decommissioned. The modelling was performed using a computer code LLWAA-DECOM (Tractebel Energy Engineering, Belgium), taking into consideration LSWS components' characteristics, parameters of the water flowing in circuits, system work regimes, etc. During the modelling, results on activity of LSWS subsystems and components' deposits and nuclide composition of the deposits at the moment of the final shutdown of the reactor, as well as activity decay of the most contaminated PCS components' deposits and dose rates after the final shutdown of the reactor were obtained.Analysis showed that there is significant difference of contamination levels among LSWS components. LSWS activity is lower than Main Circulation Circuit (MCC) and Purification and Cooling System (PCS) subsystems activity. Activity variation for LSWS components is smaller compared to MCC and PCS. Nuclides that mostly contribute to contamination in deposits of LSWS are corrosion nuclides as in MCC and PCS.

Modelling of the Radiological Contamination of the RBMK-1500 Reactor Control and Protection System Channels’ Cooling Circuit

The article presents results of modelling and analysis of component contamination of the RBMK- 1500 reactor Control and Protection System channels’ Cooling Circuit (CPSCC) at Ignalina NPP Unit 1. The modelling was performed using a computer code LLWAA-DECOM (Tractebel Energy Engineering, Belgium), taking into consideration CPSCC components characteristics, parameters of the water flowing in the circuits, system work regimes, etc. During the modelling, results on activity of CPSCC components’ deposits, nuclide composition of the deposits and dose rates after the final shutdown of the reactor, as well as activity decay of the most contaminated CPSCC components’ deposits were obtained. Analysis showed that there is a significant difference in contamination levels between CPSCC components. The rundown header from the channels of the reactor’s fast acting scram system is the most contaminated component, and contamination of the least contaminated component is only 0.27% compared to the activity of the most contaminated component. Corrosion nuclides are the nuclides that mostly contribute to contamination of the CPSCC deposits.

Actinides in irradiated graphite of RBMK-1500 reactor

The activation of graphite in the nuclear power plants is the problem of high importance related with later graphite reprocessing or disposal. The activation of actinide impurities in graphite due to their toxicity determines a particular long term risk to waste management. In this work the activation of actinides in the graphite constructions of the RBMK-1500 reactor is determined by nuclear spectrometry measurements of the irradiated graphite sample from the Ignalina NPP Unit I and by means of numerical modeling using two independent codes SCALE 6.1 (using TRITON-VI sequence) and MCNPX (v2.7 with CINDER). Both models take into account the 3D RBMK-1500 reactor core fragment with explicit graphite construction including a stack and a sleeve but with a different simplification level concerning surrounding graphite and construction of control roads. The verification of the model has been performed by comparing calculated and measured isotope ratios of actinides. Also good prediction capabilities of the actinide activation in the irradiated graphite have been found for both calculation approaches. The initial U impurity concentration in the graphite model has been adjusted taking into account the experimental results. The specific activities of actinides in the irradiated RBMK-1500 graphite constructions have been obtained and differences between numerical simulation results, different structural parts (sleeve and stack) as well as comparison with previous results (Ancius et al., 2005) have been discussed. The obtained results are important for further decommissioning process of the Ignalina NPP and other RBMK type reactors, especially for 244Cm estimation which is a critical parameter contributing to the total alpha activity of the irradiated reactor graphite for approximately 200 years.

Numerical modelling of radionuclide inventory for RBMK irradiated nuclear fuel

After the final shutdown of Ignalina NPP Unit 1 in 2004 and Unit 2 in 2009, total amount of spent nuclear fuel (SNF) is approximately 22,000 of fuel assemblies. Radionuclide content in SNF and its characteristics are initial data for analysis of various subjects such as radiation shielding, thermal analysis and other processes that can affect the performance of SNF storage or disposal facilities’ components. For instance, ageing of the components is influenced by decay heat, neutron and gamma radiation, external environment conditions, mechanical stresses, etc. The first two factors can be defined only after SNF characteristics are investigated and irradiation levels are determined. Experimental investigation of radionuclide content and characteristics in SNF is expensive and complicated, so numerical methods are widely used. In this paper, results of numerical modelling of spent RBMK-1500 nuclear fuel characteristics and its variation with time are presented. Modelling was performed using SCALE computer code system TRITON module. Some calculation results are compared with available experimental data for RBMK-1000 spent nuclear fuel and a rather good agreement between numerical and experimental results is identified. Comparison of calculated gamma and neutron sources, activities and decay heats shows that results for RBMK-1500 fuel with various enrichments differ by 10–20%.

Modelling of the Radiological Contamination of the RBMK-1500 Reactor Water Purification and Cooling System

This paper presents modelling results on the RBMK-1500 reactor water purification and cooling system (PCS) components contamination at Ignalina NPP Unit 1. The modelling was performed using a computer code LLWAA-DECOM (Tractebel Energy Engineering, Belgium), taking into consideration PCS components characteristics, parameters of the water flowing in circuits, system work regimes, and so forth. During the modelling, results on activity of PCS subsystems and components’ deposits and nuclide composition of deposits at the moment of the final shutdown of the reactor, as well as activity decay of the most contaminated PCS components’ deposits and dose rates after the final shutdown of the reactor, were obtained. Significant difference of contamination levels was revealed among PCS subsystems and subsystems components. The subsystem of nonpurified water is the most contaminated in PCS, and the activity of the least contaminated component in this subsystem is only 1.42% compared to the activity of the most contaminated component. The most contaminated and the least contaminated components of the purified water subsystem comprise 28.33% and 0.86% of activity, respectively, compared to the activity of the most contaminated PCS component.

Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight 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 slow in comparison with processes in reactor core during LOCA accident. Thus, 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 leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

Indirect assessment of 135Cs activity in the ventilation system of the Ignalina NPP RBMK-1500 reactor

Abstract The accumulation of long-lived 135 Cs isotope in the ventilation system components of the Ignalina NPP Unit 2 was investigated by spectrometric measurements and mathematical modeling. Volumetric activities of fission noble gas and other short-lived isotopes ( 41 Ar, 85m Kr, 87 Kr, 88 Kr, 88 Rb, 133 Xe, 133m Xe, 135 Xe, 138 Xe, 138 Cs) have been measured by gamma spectrometric technique. Modeling of radionuclide transport in the ventilation system provides possibility of determining essential transport parameters: effective gas flow, mean gas retention time, deposition rate of aerosols. Estimated parameters were used for indirect evaluation of difficult to measure 135 Cs isotope activity in the ventilation system components: a delay chamber and aerosol filters. The results show that the major part of 135 Cs activity is accumulated in aerosol filters, whereas the total surface activity of the delay chamber is considerably lower. Specific activities of the ventilation system components of the Ignalina NPP Unit 2 are below the clearance levels for 135 Cs.

Parameters Estimation in Modelling of Gas-Gap in RBMK Type Reactor Using Bayesian Approach

This study presents developed algorithm for assessment and updating estimates of parameters in the mathematical models of non-stationary processes (for instance, system ageing model, dynamic system models and so on) with respect of prior information and new obtained observations. Proposed algorithm for updating estimates of random parameters is based on modified application of Bayesian approach (BA). Developed algorithm was applied for Ignalina NPP Unit 2 RBMK-1500 reactor's closure of the gas-gap between the pressure tubes and the graphite bore probabilistic analysis.

Hydrogen influence on mechanical and fracture mechanics characteristics of zirconium Zr–2.5Nb alloy at ambient and elevated temperatures

This article deals with the investigation of the hydrogen concentration and temperature influence onto mechanical and fracture mechanics characteristics of RBMK-1500 Ignalina NPP unit 2 reactor fuel channel material—Zr–2.5Nb zirconium alloy (TMT-2) at temperatures from ambient up to 300 °C. The investigation of mechanical characteristics was performed on tensile specimens, fracture mechanics characteristics K Q, K C * , J IC—on compact specimens ( B = 4 mm) of hydrogen-free and saturated by hydrogen (52, 100 and 140 ppm) at 20, 170, 200 and 300 °C. The investigation showed that temperature increasing calls mechanical strength decreasing, whereas the reductions of area increase. Stronger influence of hydrogen concentration onto mechanical characteristics is noticed only at 20–170 °C temperature, however this influence diminishes as the temperature increases and weakest hydrogen influence is given at 300 °C. Fracture toughness characteristics K Q, K C * more depends on temperature than on hydrogen concentration. Critical J IC integral values for the specimens containing hydrogen were given lowest at 20 °C, increases when temperature were raised up to 140 °C and were given highest when it reaches 300 °C. The analysis of K C * and J IC dependence due to the mechanical characteristics of zirconium alloy has showed that the modified plasticity Z mod = ( R p0.2/ R m) Z satisfactorily approximates the influence of temperature and hydrogen concentration on variation of these characteristics.

SENSITIVITY OF DYNAMIC BEHAVIOUR OF THE FE MODEL: CASE STUDY FOR THE IGNALINA NPP REACTOR BUILDING

The 3D thin‐walled finite element model of Ignalina NPP Unit 2 reactor building was developed aimed at the evaluation of the global dynamic behaviour with a focus on the seismic response. The model comprises description of the monolithic structures, while prefabricated frame structures are ignored and replaced by external masses. Sensitivity study of the selected dynamic characteristics of the model with respect to data uncertainties is considered. Uncertainty of the model is considered in terms of masses of removed structures and wall stiffness. Seismic input is represented by the site specific free‐field ground response acceleration spectra. The sensitivity study concerns variations of frequencies and acceleration of in‐structure horizontal response spectra at specified points. Maximal bending moments are also considered. It was obtained that the reactor level is not sensitive to the uncertainties considered, while discernable sensitivity was detected at the top level of the structure. Santrauka Pateikta Ignalinos atominės elektrinės pastato erdvinio baigtinių elementų dinaminio modelio kūrimo koncepcija, išnagrinėtas šio modelio jautrumas keičiamoms masėms ir sienų standumui. Parodyta, kaip šie keičiami dydžiai turi įtaką dažniams, horizontaliems tam tikrų nagrinėjamų taškų atsako spektrams, lenkimo momentų persiskirstymui ir jų didžiui.

05/01038 Investigation on gasoline deep desulfurization for fuel cell applications

The paper presents modelling and analysis results of contamination of the RBMK-1500 reactor Low Salted Water System (LSWS) components and comparison of the results with other RBMK-1500 water systems at Ignalina NPP Unit 1. These results provide information for radiological characterization, which is an important task of the nuclear installation to be decommissioned. The modelling was performed using a computer code LLWAA-DECOM (Tractebel Energy Engineering, Belgium), taking into consideration LSWS components' characteristics, parameters of the water flowing in circuits, system work regimes, etc. During the modelling, results on activity of LSWS subsystems and components' deposits and nuclide composition of the deposits at the moment of the final shutdown of the reactor, as well as activity decay of the most contaminated PCS components' deposits and dose rates after the final shutdown of the reactor were obtained.Analysis showed that there is significant difference of contamination levels among LSWS components. LSWS activity is lower than Main Circulation Circuit (MCC) and Purification and Cooling System (PCS) subsystems activity. Activity variation for LSWS components is smaller compared to MCC and PCS. Nuclides that mostly contribute to contamination in deposits of LSWS are corrosion nuclides as in MCC and PCS.

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.

96/00314 Study of the structure and components of tar from Shenfu coal

After the final shutdown of Ignalina NPP Unit 1 in 2004 and Unit 2 in 2009, total amount of spent nuclear fuel (SNF) is approximately 22,000 of fuel assemblies. Radionuclide content in SNF and its characteristics are initial data for analysis of various subjects such as radiation shielding, thermal analysis and other processes that can affect the performance of SNF storage or disposal facilities’ components. For instance, ageing of the components is influenced by decay heat, neutron and gamma radiation, external environment conditions, mechanical stresses, etc. The first two factors can be defined only after SNF characteristics are investigated and irradiation levels are determined. Experimental investigation of radionuclide content and characteristics in SNF is expensive and complicated, so numerical methods are widely used. In this paper, results of numerical modelling of spent RBMK-1500 nuclear fuel characteristics and its variation with time are presented. Modelling was performed using SCALE computer code system TRITON module. Some calculation results are compared with available experimental data for RBMK-1000 spent nuclear fuel and a rather good agreement between numerical and experimental results is identified. Comparison of calculated gamma and neutron sources, activities and decay heats shows that results for RBMK-1500 fuel with various enrichments differ by 10–20%.

95/00428 The first evolutionary advanced PWR

The paper presents modelling and analysis results of contamination of the RBMK-1500 reactor Low Salted Water System (LSWS) components and comparison of the results with other RBMK-1500 water systems at Ignalina NPP Unit 1. These results provide information for radiological characterization, which is an important task of the nuclear installation to be decommissioned. The modelling was performed using a computer code LLWAA-DECOM (Tractebel Energy Engineering, Belgium), taking into consideration LSWS components' characteristics, parameters of the water flowing in circuits, system work regimes, etc. During the modelling, results on activity of LSWS subsystems and components' deposits and nuclide composition of the deposits at the moment of the final shutdown of the reactor, as well as activity decay of the most contaminated PCS components' deposits and dose rates after the final shutdown of the reactor were obtained.Analysis showed that there is significant difference of contamination levels among LSWS components. LSWS activity is lower than Main Circulation Circuit (MCC) and Purification and Cooling System (PCS) subsystems activity. Activity variation for LSWS components is smaller compared to MCC and PCS. Nuclides that mostly contribute to contamination in deposits of LSWS are corrosion nuclides as in MCC and PCS.

France's first N4 unit at Chooz B nears startup

The paper presents modelling and analysis results of contamination of the RBMK-1500 reactor Low Salted Water System (LSWS) components and comparison of the results with other RBMK-1500 water systems at Ignalina NPP Unit 1. These results provide information for radiological characterization, which is an important task of the nuclear installation to be decommissioned. The modelling was performed using a computer code LLWAA-DECOM (Tractebel Energy Engineering, Belgium), taking into consideration LSWS components' characteristics, parameters of the water flowing in circuits, system work regimes, etc. During the modelling, results on activity of LSWS subsystems and components' deposits and nuclide composition of the deposits at the moment of the final shutdown of the reactor, as well as activity decay of the most contaminated PCS components' deposits and dose rates after the final shutdown of the reactor were obtained.Analysis showed that there is significant difference of contamination levels among LSWS components. LSWS activity is lower than Main Circulation Circuit (MCC) and Purification and Cooling System (PCS) subsystems activity. Activity variation for LSWS components is smaller compared to MCC and PCS. Nuclides that mostly contribute to contamination in deposits of LSWS are corrosion nuclides as in MCC and PCS.