Ignalina NPP Research Articles

Groundwater path radiological impact assessment for the derivation of specific clearance levels for landfill disposed waste at Ignalina NPP

Industrial waste generated from the operation of Ignalina NPP was disposed of in industrial waste dumps, the so-called “landfill facility”. At the time of disposal, the waste was considered as non-radioactive. Due to changes in the regulatory requirements, a part of the waste was reclassified to a very low-level radioactive waste. The Ignalina NPP final decommissioning plan foresees the development of specific clearance levels for the disposed of waste. Application of specific clearance would allow removal from regulatory control and conversion of the existing landfill into a conventional waste disposal facility.This paper focuses on the modeling of radionuclide release and dispersion by different groundwater pathways, calculation of associated radiological impacts and definition of specific clearance levels for the landfill facility disposed of waste. The radionuclides considered in the assessment are based on the waste specific and include 3H, 60Co, 94Nb, 137Cs and 241Am. Modeling is performed using GOLDSIM software.

FE modeling of the reconstructed Bituminated radioactive waste storage of Ignalina NPP

Decommissioning of nuclear facilities is a complex of various activities strongly related to safety. In 2004 and 2005, both reactors of Ignalina NPP were closed. One of the tasks that must be solved is what to do with the radioactive waste storage facility (Ignalina AE Building 158). Possible solutions: – complete destruction of the radioactive waste storage with the need to empty the canyons and store the waste in a special place or, as an alternative, turning the storage into a repository. Converting storage to repository also has many different options. The installation of protective barriers of different thicknesses or even the construction of a sarcophagus is possible. The cheapest way to form a protective barrier on an existing building structure. The problem is the relatively high weight of the barrier structure, which must be supported by the existing building structure.The article presents an analysis of the durability of the structures of the radioactive waste storage, assessing the influence of the self-weight of the building's reinforced concrete structural elements and the mass of the bituminous radioactive waste stored in the canyons, as well as the effect of additional loads acting on the floor and roof of the first floor during the planned conversion of the storage facility into the repository. Numerical simulation was performed using ANSYS computer code and the results were compared with the relevant standards and norms in order to assess the building's ability to undergo such transformations.Spatial volumetric geometric and calculation models of the construction structures of the 158 building were created. These models have been modified after considering future changes during the transformation of the building into a repository. Perform calculations of the state of stresses and strains of the landfill under the load of the self-weight of the structural elements and the hydrostatic pressure of the bituminous waste. On the basis of the results of the calculations, conclusions were formulated about the level of load on the construction structures of the building and the margin under the current operating loads.The developed numerical models can be easily extended to evaluate the effects of new building elements and additional loads on structures (specifically walls and ceilings) after the transformation of a waste storage into a repository.The article presents not only the idea of reorganizing the waste storage structure to repository but also the methodology of modelling such a structure by assessing the age of the building's soil and additional loads. This methodology can be useful and applied to the closure of other nuclear power plants when it is necessary to turn the radioactive waste storage into a repository.

Analysis of loss of cooling accidents in the spent fuel pools using system thermal hydraulics and computational fluid dynamic codes

The interest in the safety of on-site storage of Spent Fuel Assemblies (SFAs) has been increased after the accident on 11 March 2011, in Fukushima Daiichi Nuclear Power Plant. It is important to analyse the behaviour of not only integral effects in spent fuel pools, but also the local effects during heat removal from the SFAs. Usually, for the integral analysis the one-dimensional system computer codes were used. These codes allow to analyse whole pools and cover long time, with a relatively low cost of computational resources. The Computational Fluid Dynamic (CFD) codes could perform very detailed analysis, but this requires large computational resources. The strong side of CFD codes is the possibility to assess local effects which could occur.In the current work the analysis of heat transfer from the SFAs, placed in special baskets was performed using ANSYS FLUENT 17.2 and RELAP5 codes. The models of single basket with SFAs were developed using both codes. The comparison of calculations allowed to evaluate capabilities of both codes. It is shown, that using one-dimensional system thermal hydraulics code developed model allows to describe coolant mixing inside basket sections and received results fits well with CFD code calculation results. Based on gained experience, the loss of cooling accident in the spent fuel pools of Ignalina NPP was analysed using the model, developed by employing RELAP5 (one-dimensional system thermal hydraulics) code. The calculation results were used for safety evaluation of Ignalina NPP facilities.

Radiocarbon generation and atmospheric release assessment from nuclear power plant with RBMK type reactors

The article provides the results of the estimation of 14C nuclide generation and release from the systems and structures of RBMK-1500 reactors operated at Ignalina NPP. The main circulation circuit and gaseous circuit of this type of reactor are the primary sources of 14C generation and airborne release, evaluated to be 0.491 and 0.367 TBq per one calendar year (10.5 months of effective reactor operation at 1350 MWe power), respectively. The analysis of the operation specifics of these systems showed that they don’t provide an effective barrier to contain 14C due to the 14CO2 chemical form of radiocarbon in RBMK reactors, i.e., 100% release (0.858 TBq/year) have been evaluated available for atmospheric release. These results have been compared with the atmospheric releases evaluated from the 14C activity concentration measured in Scots pine tree (Pinus Sylvestris L.) ring samples by single stage accelerator mass spectrometer (SSAMS, NEC, USA) at Vilnius Radiocarbon facility, resulting to airborne release value of 1 TBq/year for one RBMK-1500 reactor. Slightly higher value reproduced from the tree ring samples indicates that the reactor graphite stack might be also a source of 14C releases of about 0.14–0.24 TBq/year. Based on similar design features the results of the study can be easily projected to the RBMK-1000 reactors rescaling 14C source term according to the lower reactor power.

A proposed approach for the evaluation of consequences of a large aircraft crash accident at an RBMK type reactor site during decommissioning

Nuclear power plants with RBMK reactors were designed in late sixties - early seventies and were built entirely in the European part of the Russian Federation, the Lithuanian Republic and the Ukrainian Republic. The plants in the Lithuanian and the Ukrainian Republics are in various stages of decommissioning. As currently planned, the most of RBMK reactors still operating in the Russian Federation will end their operation in this decade.The reactor building, spent nuclear fuel and radioactive waste storage facilities, built at the time of construction of RBMK plants are not specially designed to withstand the direct impact of an aircraft crash. There were no such regulatory requirements at the time the plants were designed and, no such requirements were introduced by regulatory bodies later on. But for now, in the aftermath of the September 11, 2001 terrorist attack on the United States, the event of an aircraft crash on a nuclear facility can no longer be considered as extremely unlikely.This paper considers the state of a nuclear power plant with the RBMK type reactor(s) during certain decommissioning and dismantling stages and discusses the approach for the evaluation of radiological consequences to the public in the event of a large aircraft crash on potentially hazardous objects of the plant. The damage to spent nuclear storage facilities, the reactor graphite stack and storage facilities for the combustible waste may cause a significant release of radionuclides to the environment. Key components of the approach include the evaluation of a structural damage to the facilities, the evaluation of a fire propagation and impact, the evaluation of a spent nuclear fuel cooling capability in storage pools, the evaluation of an airborne radionuclide release, the evaluation of a radionuclide dispersion in the environment and the evaluation of the public exposure. A set of widely used software and methods for the performance of the specific tasks is discussed. The paper also describes how the proposed approach was applied for the evaluation of radiological consequences from a Boeing 767-400 type aircraft crash at the Ignalina NPP site where a plant with two RBMK-1500 type reactors is decommissioned.The paper does not address modern radioactive waste and spent nuclear fuel management and storage facilities that might be built at RBMK sites as well. As a rule, assessment and mitigation of an aircraft crash accident consequences is an integral part of design of these new facilities.

A review of uncertainties in the assessment of the radiological impact of liquid releases to the Ignalina NPP cooling pond lake Drūkšiai

The methodology used for assessment of the public exposure from the Ignalina nuclear power plant (INPP) liquid radioactive releases to the plant cooling pond, Lake Drūkšiai is based on the application of so called “release-to-dose calculation factors” (RDFs). The application of RDFs is straightforward and was used to assess impacts of actual and planned releases. The RDFs were derived a few decades ago, and thus are based on that time understanding of the plant operation and the environment characteristics. The early shutdown of the INPP power reactors and the startup of the plant decommissioning changed the initially planned conditions for radionuclide release and dispersion. The appropriateness of RDF values became questionable in their past application for prospective and in their current application for retrospective impact assessments.This publication addresses uncertainties in RDF values taking into consideration the actual operation and the planned decommissioning of the INPP. The model for calculation of RDFs for liquid releases was updated to the current status of knowledge and regulatory requirements without changing the complexity of the model itself. The analysis includes a maximum possible or reasonably large variation of the model parameters, defining radionuclides dispersion in the environment and exposure to the population.The assessment results show that RDF values for majority of the considered radionuclides 3H, 54Mn, 55Fe, 60Co, 90Sr, 134Cs and 137Cs are supported by the probabilistic analysis of the model, scenario and parameter uncertainties. The RDF values for radionuclides 93mNb and 94Nb are overestimated, while a higher RDF value for radionuclide 14C is recommended.

Scientific inheritance of Eugenijus Ušpuras

Abstract This article is written to honour the scientific achievements of prof. Eugenijus Ušpuras, who passed away in 2019. Therefore, his colleagues summarized retrospectively main research topics like safety studies for Ignalina NPP, research in nuclear fusion and investigations in energy security prof. E. Ušpuras worked with. Since prof. Eugenijus Ušpuras spent his entire scientific career at Lithuanian Energy Institute (LEI) all these three topics are directly related to the development of the nuclear energy topics and research performed in Lithuania during the last decades.

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 Contamination by 14C Airborne Releases in the Vicinity of the Ignalina Nuclear Power Plant

ABSTRACTA radiocarbon (14C) activity analysis in the tree rings around Ignalina nuclear power plant (INPP) has been carried out with the aim to test the hypothesis to use 14C tree-ring analysis data as a tool for the reconstruction of gaseous releases from NPP to the environment. The INPP has been in decommissioning state since the end of 2009. Tree-ring samples for 14C analysis were collected 7 yr after final power unit shutdown from the INPP vicinity. The samples from 5 sampling locations were collected, prepared and measured using the Single Stage Accelerator Mass Spectrometer (SSAMS). Data analysis represents observable Ignalina NPP influence by 14C increase up to 15 pMC (percent modern carbon) in tree rings. Good correlations of the 14C concentrations and wind direction were obtained. The main purpose of this article was to match 14C measurement data along with the atmospheric dispersion modeling of emissions in order to retrospectively characterize the emission source.

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).

Disposal of very low-level radioactive waste: Lithuanian case on the approach and long-term safety aspects

Decommissioning of nuclear power plants typically requires management of hundreds of thousands of cubic meters of waste with very low levels of radioactivity. Such very low-level waste (VLLW) is suitable for disposal in simple landfill type facilities.The only nuclear power plant in the Republic of Lithuania, the Ignalina NPP, was finally shut down at the end of 2009. Final management of the existing operational radioactive waste and that arising during the decommissioning will require the disposal of 60,000 m3 of very low-level short-lived solid waste. An overview of the implementation of the landfill disposal facility for very low-level radioactive waste at the Ignalina NPP site and the main aspects of long-term safety assessments based on national regulations as well as the IAEA's recommendations including the siting process are presented in this paper as well as a short description of the disposal concept and the characteristics of the waste intended for disposal in the landfill. Water pathway scenarios as well as intrusion scenarios have been investigated following an IAEA methodology. It has been defined that 14C is the greatest contributor to the total dose in the case of water pathway scenarios. In the case of intrusion scenarios, the main contribution to the total dose is from radionuclides 94Nb and 137Cs. In all the considered cases, the total value of the annual effective exposure dose remains quite well (orders of magnitude) below the value of the dose constraint established by regulations of Lithuania. Uncertainty analysis is carried out to define the influence of various factors on the obtained results of the safety assessment.

Preliminary investigation of14C migration from RBMK-1500 reactor graphite disposed of in a potential geological repository in crystalline rocks in Lithuania

ABSTRACTThere are two units with RBMK-1500 type reactors at the Ignalina Nuclear Power Plant (Ignalina NPP) in Lithuania where graphite was used as a neutron moderator and reflector. These reactors are now being decommissioned, and Lithuania has to find a solution for safe irradiated graphite disposal. It cannot be disposed of in a near surface repository due to large amounts of14C (radiocarbon, carbon-14); thus, a deep geological repository (DGR) is analyzed as an option. This study had the aim to evaluate14C migration from the RBMK-1500 irradiated graphite disposed of in a potential DGR in crystalline rocks taking into account the outcomes of the research performed under the collaborative European project CAST (CArbon-14 Source Term) and to identify the potential to reduce the conservatism in the assumptions that was introduced in the lack of data and led in the overestimated14C migration. The information gathered during the CAST project was used to model14C transport in the near field by the water pathway and to perform uncertainty analysis. The study demonstrated that more realistic assumptions could reduce the estimated14C flux from the near field by approximately one order of magnitude in comparison with the previous estimations based on very conservative assumptions.

Estimation of the radiation doses during the dismantling of the equipment in Building 117/1 at the Ignalina NPP

The Ignalina NPP was operating two RBMK-1500 reactors which are under decommissioning now. Building 117/1 was the first dismantling project at the Ignalina NPP. The DECRAD computer code was used to plan dismantling and decontamination (D&D) activities in Building 117/1. The D&D activities in Building 117/1 were finished at the end of 2011. In this paper, collective doses, duration of the dismantling activities and other parameters defined during the planning of the D&D activities for Building 117/1 are compared with the real dismantling data and also with updated modelling results.

Tritium in the environment of the Ignalina NPP in the period of its operation

Tritium in the environment of the Ignalina NPP in the period of its operation

The distribution of tritium in aquatic environments, Lithuania

The aim of this study is to investigate mobile radionuclide tritium (3H or T) activity dynamics in aquatic environments related to Ignalina NPP (INPP) site and water bodies located in remote areas unaffected by the INPP. The 3H excess in the INPP environment was analyzed and compared to the variable 3H background level over the period of operation of the INPP (end of 1983 – end of 2009) and during the initial stage of decommissioning (2010–2017). 3H in the INPP vicinity has been studied in the water of artificial channels related to operation of the INPP and site drainage, in natural surface water bodies and, at a smaller scale, in unconfined groundwater. This study presents an extensive 3H data set extending back to 1980, i.e. before INPP operation started. To assess the contribution of global sources to 3H dynamics, monthly precipitation was also studied, along with water from the Baltic Sea, Curonian Lagoon and Nemunas River were studied as well, all three of these located in the Lithuanian maritime zone. The 3H activity concentration in water was measured using liquid scintillation counting (LSC) techniques (direct counting and counting after enrichment). During the period of INPP operation, 3H from liquid effluent could be clearly observed in discharge channels, occurring in rather low diluted conditions, as well as in Lake Druksiai, the cooling basin, at an even more diluted level. The highest 3H activity concentration in Lake Druksiai was observed in 2003 and reached 201.3 ± 1.3 TU at a time when 3H activity concentrations in background water bodies was 9.2 ± 3.5 TU. After the closure of the INPP, the 3H liquid effluent rate reduced by approximately two orders of magnitude (from 1012 Bq in 1991 to 1010 Bq in 2016) and when decommissioning activity commenced then the 3H activity concentration fell to that approaching the background level (19–27 TU) that can still be observed in industrial discharge and rainwater drainage channels. 3H as a result of leakage from the INPP can be observed in groundwater only in direct proximity to the INPP site near the radioactive waste storage zone.

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.

Assessment of Political Vulnerabilities on Security of Energy Supply in the Baltic States

Abstract The article argues that despite the evident link between political environment and security of energy supply, political elements are not sufficiently represented in contemporary scientific literature, namely in indexes that are designed for the assessment of security of energy supply. In an attempt to fill this gap, the article presents an innovative methodology for quantitative assessment of the political vulnerabilities on security of energy supply and applies it to the analysis of the Baltic States. The proposed index determines the plausibility of the occurrence of threats of a political nature on the security of energy supply and defines it as political vulnerability. The application of index methodology to an analysis of the Baltic States has revealed that the overall political vulnerability on security of energy supply is the highest in Lithuania, considerably lower in Latvia, and the lowest in Estonia. The analysis has shown that political vulnerability has increased in Lithuania due to the closure of Ignalina NPP and an increase in energy import quantities from politically unstable countries, such as Russia. On the contrary, political vulnerabilities on the security of the energy supply have decreased in Latvia and Estonia due to the increase of consumption of indigenous energy. However, preliminary calculations show that political vulnerabilities should decrease considerably in 2015 in Lithuania due to the diversification of the natural gas supply.

HAZOP application for the nuclear power plants decommissioning projects

Decommissioning of nuclear facilities involves different types of activities, tools, equipment and systems. There is a potential for a wide range of radiological and industrial accidents during various stages of a decommissioning project creating risk for workers and the environment. The occurrence of accidents is possible due to many different operations involving movement and handling of large pieces of equipment and contaminated items. In addition, size reduction and decontamination processes are capable of producing hazards. One of the first steps in developing a safety assessment for decommissioning activities is the identification of hazards that can affect workers, members of the public and the environment during decommissioning activities, and then to identify engineered and administrative control measures to prevent, eliminate or mitigate the hazards and their consequences. Fault and hazard identification can be undertaken in several ways using a range of tools and techniques, including Hazard and Operability Study (HAZOP).The paper will mainly focus on the application of HAZOP technique for identification of the hazards raised due to dismantling and decontamination activities at the Ignalina NPP, as well as at feasibility study for the management of Bohunice V1 NPP primary circuit components.

Updating of the public dose assessment approach for decommissioning related releases from the Ignalina NPP

The methodology used at the Ignalina NPP for the assessment of the public exposure from radioactive releases to the environment is based on the use of so called “release-to-dose calculation factors” (RDFs). Application of RDFs is straightforward and in practice is used both to assess impacts from measured releases and to assess impacts from planned activities. The set of radionuclides and RDF values corresponding to them for operational releases were approved by the regulatory authorities.The decommissioning-related radiological characterization of the plant components and wastes revealed that the available set of RDFs is not sufficient for the decommissioning needs. To overcome this, a simplified approach has been proposed for the estimation of the missing RDF values. However, estimations using the simplified approach might be very inaccurate and not necessary conservative.This paper addresses the current practice of the dose assessment for releases from the Ignalina NPP and the necessity of extension of the available set of RDFs. New RDF values for the calculation of radiological impacts due to airborne releases from the power units and waterborne releases into Lake Drūkšiai are proposed for the decommissioning-relevant radionuclides. The significance of individual radionuclides in releases to the environment during decommissioning is discussed.

The assessment of ionising radiation impact on the cooling pond freshwater ecosystem non-human biota from the Ignalina NPP operation beginning to shut down and initial decommissioning

The radiological doses to non-human biota of freshwater ecosystem in the Ignalina NPP cooling pond – Lake Druksiai were evaluated for several cases including the plant's operation period and initial decommissioning activities, using the ERICA 1.2 code with IAEA SRS-19 models integrated approach and tool.Among the Lake Druksiai freshwater ecosystem reference organisms investigated the highest exposure dose rate was determined for bottom fauna – benthic organisms (mollusc-bivalves, crustaceans, mollusc-gastropods, insect larvae), and among the other reference organisms – for vascular plants. The mean and maximum total dose rate values due to anthropogenic radionuclide ionising radiation impact in all investigated cases were lower than the ERICA screening dose rate value of 10 μGy/h. The main exposure of reference organisms as a result of Ignalina NPP former effluent to Lake Druksiai is due to ionizing radiation of radionuclides 60Co and 137Cs, of predicted releases to Lake Druksiai during initial decommissioning period – due to radionuclides 60Co, 134Cs and 137Cs, and as a result of predicted releases to Lake Druksiai from low- and intermediate-level short-lived radioactive waste disposal site in 30–100 year period – due to radionuclides 99Tc and 3H.The risk quotient expected values in all investigated cases were <1, and therefore the risk to non-human biota can be considered negligible with the exception of a conservative risk quotient for insect larvae. Radiological protection of non-human biota in Lake Druksiai, the Ignalina NPP cooling pond, is both feasible and acceptable.