RBMK Type Research Articles

What cooling pond sediments can reveal about 14C in nuclear power plant liquid effluents: Case study Lake Drūkšiai, Ignalina nuclear power plant cooling pond.

The vertical distribution of radiocarbon (14C) was examined in two bottom sediment cores, taken from Lake Drūkšiai which had been used as a cooling pond for the Ignalina nuclear power plant (INPP) with two RBMK type reactors. The aim of this work was to reconstruct 14C amounts in the lake ecosystem during an 8-year period after the INPP was closed, as any official monitoring of 14C in liquid releases from the INPP was not performed. The possibility of comparing the variation of the 14C specific activity in the corresponding layers of the same period of 3 different cores (one taken in 2013 and two in 2019) revealed the variability of the determined values of liquid radiocarbon discharges from the INPP into the lake. Cores taken in 2019 showed a permament14C release rate of 0.76±0.06 GBq/y all eight years after the closure of the INPP. The 14C release rate established from radiocarbon measurements in both cores did not differ by more than 0.8 GBq/y. However, including data from the core taken several years ago, the estimated radiocarbon release rate values varied within 1.3 GBq/y.

Predictive diagnostics of RBMK reactor flowmeters

At the moment, there are 8 rectors of the RBMK type in operation. The safety of their operation is ensured by monitoring devices for key indicators. In this article, the main attention is paid to the control of the coolant flow through the fuel channel, carried out with the help of ball flow meters of the STORM-32M and STORM-8A types. During operation, there is a gradual wear of the raceway and the ball, which leads to the appearance of a negative error in the device, in other words, the readings become underestimated. In this regard, as well as due to the widespread use of ball flowmeters at nuclear power plants, the task of diagnosing the state of flowmeters is urgent. One of the approaches to determining the flow rate of the coolant is an approach based on the measurement and use of information about the activity of the coolant by the tightness control system of the fuel element shells (SKGO). The use of this radiation approach to control the flow of coolant has its advantages and difficulties. The advantage is that real experimental information about the presence of a coolant in the fuel channel is involved, and the difficulties are associated with receiving and processing the signals of the SCGO in real time and, consequently, with low efficiency of solving the problem. In this paper, we propose a method for predictive diagnostics of the flow meter condition using the design features of the multiple forced circulation circuit (CMPC) of the reactor. The adaptive parameter of the model was selected as a diagnostic parameter. Studies have shown that the increase in time variance in experimental flow values has a clear effect on the selected diagnostic feature , as a result of which it can be said that the observation of this kind of variance change in experimental data can serve as a sign of a malfunction of the flow meter.

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.

Modeling radionuclide migration from activated metallic waste disposal in a generic geological repository in Lithuania

There are two RBMK-1500 type reactors in Lithuania, which are under decommissioning now. Long-lived low and intermediate level radioactive metallic waste generated during operation and decommissioning of the nuclear power plant is foreseen for deep geological disposal (DGR). A preliminary assessment of the radionuclide release from metallic waste and migration through the engineered barriers of the generic DGR in crystalline rocks is presented in this study. The radionuclide flux to backfill and geosphere is investigated for three different container alternatives: (1) disposal of metallic waste in stainless steel containers, (2) disposal in carbon steel containers and (3) disposal in which the container is not a barrier for radionuclide migration. In addition, the impact of surface contamination of the activated waste on the radionuclide flux is investigated. Comparison of maximum fractional flux to backfill between Alternatives 1 and 2 revealed that the container material plays a noticeable role in the migration of long-lived low mobility radionuclides. The maximum fractional flux to the geosphere for all radionuclides increases from Alternative 1 to Alternative 3. Surface contamination is an important parameter when the maximum radionuclide flux to backfill is observed during the early post-closure period.

Distribution of radiocarbon in sediments of the cooling pond of RBMK type Ignalina Nuclear Power Plant in Lithuania.

The vertical distribution of radiocarbon (14C) was examined in the bottom sediment core, taken from Lake Drūkšiai, which has served as a cooling pond since 1983 for the 26 years of the Ignalina Nuclear Power Plant (INPP) operation using two RBMK-1500 reactors (Russian acronym for”Channelized Large Power Reactor”). 14C specific activity was measured in alkali-soluble and -insoluble fractions of the sediment layers. Complementary measurements of the 210Pb and 137Cs activity of the samples provided the possibility to evaluate the date of every layer formation, covering the 1947–2013 period. In addition, 14C distribution was examined in the scales of pelagic fish caught between 1980 and 2012. Our measurements reveal that, during the period 1947–1999, the radiocarbon specific activity in both fractions exhibits a parallel course with a difference of 5 ± 1 pMC (percent of modern carbon) being higher in alkali-soluble fraction, although 14C specific activity in both fractions increased by 11.4–13.6 pMC during the first 15 years of plant operation. However, during the 2000–2009 period, other than previously seen, a dissolved inorganic carbon (DIC) → aquatic primary producers → sediments 14C incorporation pattern occurred, as the radiocarbon specific activity difference between alkali-soluble and -insoluble fractions reached 94, 25, and 20 pMC in 2000, 2006, and 2008, respectively. Measurements in different sediment fractions allowed us to identify the unexpected organic nature of 14C contained in liquid effluences from the INPP in 2000–2009. The discrepancy between 14C specific activity in fish scales samples and DIC after 2000 also confirmed the possibility of organic 14C contamination. Possible reasons for this phenomenon might be industrial processes introduced at the INPP, such as the start of operation of the cementation facility for spent ion exchange resins, decontamination procedures, and various maintenance activities of reactor aging systems and equipment.

Evaluation of coolant density history effect in RBMK type fuel modelling

The axial heterogeneous void distribution in a fuel channel is a relevant and important issue during nuclear reactor analysis for LWR, especially for boiling water channel-type reactors. Variation of the coolant density in fuel channel has an effect on the neutron spectrum that will in turn have an impact on the values of absolute reactivity, the void reactivity coefficient, and the fuel isotopic compositions during irradiation. This effect is referring to as the history effect in light water reactor calculations. As the void reactivity effect is positive in RBMK type reactors, the underestimation of water density heterogeneity in 3D reactor core numerical calculations could cause an uncertainty during assessment of safe operation of nuclear reactor. Thus, this issue is analysed with different cross-section libraries which were generated with WIMS8 code at different reference water densities. The libraries were applied in single fuel model of the nodal code of QUABOX-CUBBOX/HYCA. The thermohydraulic part of HYCA allowed to simulate axial water distribution along fuel assembly model and to estimate water density history effect for RBMK type fuel.

Heat removal from RBMK reactor core using non-regular means

RBMK type reactor is a boiling light-water-reactor with a graphite moderator. Several important design features of RBMK are unique and very complex compared with: the fuel assemblies are located in individual channels rather than a single pressure vessel, these fuel channels are placed in a massive graphite moderator block; the rods of reactor control and protection system are placed in different channels, parallel to the fuel channels in graphite moderator block. Such design provides additional opportunity for heat removal from the reactor in the cases of beyond design-basis accidents. This paper discusses possibility of heat removal from the RBMK-1500 core using weak heat conduction mechanisms: heat removal from outer surface of drum separators and steam – water piping using air ventilation means; heat removal from graphite stack using direct water supply into reactor cavity; heat removal from fuel channels through the core graphite structures by control rods cooling system; heat removal in loss of long term cooling case using de-pressurisation of the reactor coolant system and water supply using complex of non-regular means. The provided analyses discuss advantages and capacities of these different means.

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).

Simplified Mathematical Model and Realization of the Remaining Life of Nuclear Fuel Upon Decommissioning of RBMK-Type Reactor

The realization of the residual life of nuclear fuel upon decommissioning of a RBMK type channel reactor is studied. A mathematical model of the energy-production spectrum of FA is presented. Two methods of decommissioning a reactor depending on the method of criticality conservation are compared. The first possible method of conserving the criticality of a reactor upon commissioning is extraction of the FA with high energy production and the second is formation of a two-zone load: fresh FA are placed at the center and spent FA at the periphery. The second method gives a greater advantage in terms of power and remaining energy-production life. The results were obtained in a point model of the dynamics of the FA spectrum in a linear fit to the FA power and multiplication coefficient versus energy production and attest the fundamental possibility of optimizing fuel use and decommissioning of a RBMK reactor.

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.

Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (keff) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15GWd/tU burn-up for 2.8% enrichment fuel and from 9GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality safety.

Modeling of irradiated graphite 14C transfer through engineered barriers of a generic geological repository in crystalline rocks

There are two RBMK-1500 type graphite moderated reactors at the Ignalina nuclear power plant in Lithuania, and they are under decommissioning now. The graphite cannot be disposed of in a near surface repository, because of large amounts of 14C. Therefore, disposal of the graphite in a geological repository is a reasonable solution. This study presents evaluation of the 14C transfer by the groundwater pathway into the geosphere from the irradiated graphite in a generic geological repository in crystalline rocks and demonstration of the role of the different components of the engineered barrier system by performing local sensitivity analysis. The speciation of the released 14C into organic and inorganic compounds as well as the most recent information on 14C source term was taken into account. Two alternatives were considered in the analysis: disposal of graphite in containers with encapsulant and without it. It was evaluated that the maximal fractional flux of inorganic 14C into the geosphere can vary from 10−11y−1 (for non-encapsulated graphite) to 10−12y−1 (for encapsulated graphite) while of organic 14C it was about 10−3y−1 of its inventory. Such difference demonstrates that investigations on the 14C inventory and chemical form in which it is released are especially important. The parameter with the highest influence on the maximal flux into the geosphere for inorganic 14C transfer was the sorption coefficient in the backfill and for organic 14C transfer – the backfill hydraulic conductivity.

Overview of the Lithuanian programme for disposal of RBMK-1500 spent nuclear fuel

AbstractIn Lithuania all the spent nuclear fuel (SNF) came from operation of the Ignalina nuclear power plant with two reactors of RBMK type (RBMK is a Russian acronym for 'Channelized Large Power Reactor' which is a water-cooled graphite-moderated reactor: RBMK-1500). Approximately 22,000 SNF assemblies are due for geological disposal in Lithuania. Currently it is envisaged that SNF will be stored in dry interim storage facilities (new and existing) for at least 50 y prior to possible deep geological disposal.The decision on the final SNF management option (disposal in a national repository, disposal in regional repositories, etc.) has not yet been made but some investigations of the possibilities to dispose of the SNF in Lithuania have been initiated. With the support of Swedish experts, analysis of possible geological formations for SNF disposal was performed and the existence of potentially suitable formations agreed. The geological formations prioritized as prospective include the crystalline rocks in southern Lithuania and two clayey formations: the Lower Triassic clay formation and the Lower Cambrian Baltic Group clay formation, with priority given to the Lower Triassic clay formation.This paper presents the main aspects of the research and other activities undertaken over the past decade in the field of SNF disposal: international cooperation; current status and plans for the Lithuanian national program; further investigations required; and competence developments.

Methodology for the scoping assessment of radioactive releases during dismantling of RBMK-1500 reactor systems with low-level contamination

The Ignalina nuclear power plant operated two reactors of the RBMK-1500 type which are now under decommissioning. Decommissioning has been started from the reactor’s periphery, with dismantling of non-contaminated and low contaminated equipment and installations. This paper discusses a method for the scoping assessment of airborne radioactive releases to the environment when the number of items of equipment and the variety of installations to be addressed is significant, and separate consideration of each item becomes impractical, especially when the level of radioactive contamination is low. The method is based on the grouping of items into the manageable number of mass streams with particular geometric and other decommissioning-approach-related properties. Radioactive releases are then evaluated by application of “mass factors” that are detailed for two geometric forms – pipes and plates. These forms are the most usual outcomes from Ignalina nuclear power plant decommissioning activities performed on reactor periphery components. The method can be used for the scoping assessment of radionuclide releases considering that the radiological impact is expected to be low and, therefore, simple and straightforward solutions are sufficient for demonstration of compliance with the established radiation protection goals.

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.

Thermal analysis of reference repository for RBMK-1500 spent nuclear fuel in crystalline rocks

Total amount of RBMK-1500 type spent nuclear fuel (SNF) in Lithuania is approximately 22 thousands of fuel assemblies. All these assemblies should be stored for 50 years and then disposed of. International consensus prevails that SNF and long-lived high-level radioactive wastes are best disposed of in geological repositories using a system of engineered and natural barriers. Disposed nuclear waste induces a number of coupled thermo-hydro-mechanical processes around the repository. Thermal analysis of a deep geological repository could provide temperature distribution which is required for the repository’s design and for evaluation of thermal integrity of the engineered barriers. One of the most critical parameters for the repository is peak temperature at the outer surface of the canister. This temperature cannot exceed 100 °C; otherwise, unfavourable groundwater chemistry can adversely affect chemical stability of the engineered barriers. Thermal behaviour of the conceptual Lithuanian repository for RBMK-1500 type SNF in crystalline rocks was modelled using numerical codes ANSYS FLUENT and COMPASS. Very similar temperature distributions around the disposal canister were determined using both modelling tools. The modelling results revealed the importance of coupled heat and hydrodynamical processes for peak temperature in the engineered barriers, whereas the impact of mechanical processes evaluation was insignificant. It was also determined that peak temperature at the outer surface of the disposal canister does not exceed the permitted 100 °C.

Full-scale mathematical model for describing the temperature conditions of a reactor of the RBMK type in the case of an accident with complete deenergization

This paper describes a full-scale mathematical model used for the analysis of temperature conditions of an RBMK nuclear reactor in the event of its complete deenergization. The results of the verification and some computational results are presented.

Overview of safety improvement during RBMK-1500 reactor core lifetime upgrading

Abstract Starting from early to mid 1970s the first generation of channel-type graphite-moderated boiling water reactors socalled RBMK types were designed and went into operation. Nowadays, all designed RBMKs represent three generations of reactors having significant differences with respect to their safety design features. The Ignalina NPP with two RBMK-1500 units being a second generation type were originally the most powerful reactors in the world. The change of important neutron-physical characteristics, being specific safety features for RBMK type reactors during RBMK-1500 core upgrading processes are analyzed until final shutdown and their impact on safety during reactor lifetime is discussed.

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.