RBMK Reactors Research Articles

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.

Special monitoring results for determination of radionuclide composition of Russian NPP atmospheric releases

Abstract Measurements of activity concentrations of radionuclides in atmospheric releases were performed in 2017–2018 at vent stacks of seven Russian nuclear power plants. The selected instruments and research methods, with detection limits significantly lower than the existing detection limit of Russian NPPs routine control, allowed to reliably determine up to 26 radionuclides. Analysis of experimental data allows to determine the list of radionuclides for calculation the effective dose rates to public and the permissible annual discharge levels for each Russian NPP. Radiocarbon is determined as major contributor for the dose from the atmospheric releases of LWGR reactors – up to 98% for EGP-6 and RBMK-1000 (Smolensk NPP) reactors. For PWR reactors (VVER) radionuclides contribution to the annual dose from atmospheric releases is more complicated, but, in general, dose is formed by tritium, 14C and noble gases. The special monitoring results with ranking of measured radionuclides according to their contribution to the effective dose makes it possible to optimize the list of controlled radionuclides in airborne releases of Russian NPPs from 94 to 8–16 for different NPPs.

Structural investigation of RBMK nuclear graphite modified by 12C+ ion implantation and thermal treatment

Graphite is a major core component in the graphite-moderated RBMK-1500 nuclear reactor. Neutron irradiation induces radiation damage in the graphite matrix, which affects physical properties of material. In order to generate defects that could be similar to those induced by neutrons in the RBMK-1500 nuclear reactor, the raw RBMK graphite samples were implanted with 12C+ ions at the energy of 700 keV at varying fluences ranging from 1.2 × 1015 to 1.2 × 1016 ions/cm2. The ion beam interaction simulations were performed by using SRIM-2013 and GEANT4 codes. In order to study temperature effects on the evolution of the ion implanted graphite structure the thermal treatment was applied. The structural disorder level on the surface of RBMK graphite samples was investigated by using Raman spectroscopy before and after ion implantation as well as after thermal treatment. The decrease of the graphitic sp2-related content as well as formation of an amorphous structure were observed on the surface of the ion implanted samples, while thermal treatment resulted in significant reordering at the ion implantation fluence of 7.2 × 1015 ions/cm2 and lower. The understanding of structural dynamics of nuclear graphite under irradiation conditions provides insights into migration of impurities (e.g. 14N) as well as distribution of activation products (e.g. 14C, 36Cl) in the irradiated graphite matrix.

Modeling of decay heat removal from CONSTOR RBMK-1500 casks during long-term storage of spent nuclear fuel

Management of spent nuclear fuel is a very important part in the whole cycle of nuclear energy generation. Ignalina nuclear power plant operated two RBMK-1500 reactors that are now being decommissioned. After careful consideration “dry” storage technology in casks was selected for the interim storage of spent nuclear fuel (SNF) for up to 50 years. SNF after pre-storage time in water pools for not less than 5 years is loaded into cast-iron or metal-concrete casks. In this paper decay heat removal was modelled from specific GNB (Gesellschaft für Nuklear Behälter GmbH) metal-concrete (CONSTOR RBMK-1500) casks for the long-term storage period up to 300 years. The ALGOR code was used for the numerical modeling of the distribution of the heat fluxes and temperatures in a loaded cask placed in an open type storage facility in both summer and winter taking into account local environmental conditions. A local sensitivity analysis of the impact of fuel parameter uncertainties is also performed.

Integrity control of an RBMK-1500 fuel rod locally oxidized under a bounding reactivity-initiated accident

In 2007, the license for the second reactor unit of the Ignalina nuclear power plant was renewed considering the safety-related modifications introduced in this reactor. The Safety Analysis Report for this reactor unit was prepared with more strict criteria. The bounding reactivity-initiated accident (RIA) performed by the Lithuanian Energy Institute could be mentioned as an example. The performed analysis demonstrated that even when the worst initial conditions and possible uncertainties are considered, the fuel cladding remains intact. However, the analysis was performed assuming a fresh fuel assembly. In this study, an analysis of the fuel rod cladding behavior in the RBMK-1500 reactor following a bounding RIA is performed using the computational codes FEMAXI-6 and RELAP5. The analysis is extended by modeling an oxide layer (nodular corrosion) on the external surface cladding. An uncertainty and sensitivity analysis was performed using a method developed by the Society for Plant and Reactor Safety, employing the Software for Uncertainty and Sensitivity Analyses, in order to evaluate the effect of the oxide layer on the inside and outside fuel rod temperatures. The results of the thermo-mechanical analysis (stress, strain, and enthalpy) for a local oxide layer with a thickness of 70 μm show that despite the exceeded limit of allowed linear power density, the fuel rod is under acceptable safety conditions.

Radiation degradation mechanisms of reactor graphites properties

Reactor graphite is used as the material for the graphite stack of Russian RBMK power reactors. Long-term operation of the reactor’s graphite stack at high operating temperatures (500-700)°С and neutron irradiation leads to a significant degradation of the graphite’s properties, which may limit the lifetime of the entire nuclear power plant. In this paper the properties degradation mechanisms of graphite materials at different stages of RBMK reactors operation was studied to justify their service life.

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.

Investigation of Impurities of RBMK Graphite by Different Methods

ABSTRACTSamples of graphite from a RBMK-1500 reactor at the Ignalina Nuclear Power Plant from different construction elements (stack, sleeve, and bushing) were analyzed by the instrumental neutron activation analysis (INAA) method (LVR-15 experimental reactor of the Research Centre Řež, Ltd.) using the prompt gamma activation analysis (PGAA) method (Heinz Maier-Leibnitz Zentrum) and with an inductively coupled plasma mass spectrometer (ICP-MS) (CPST, Lithuania). These measurements were performed with the aim of obtaining the missing information on the impurity distribution in the RBMK-type nuclear graphite constructions as well as for intercomparison purposes, with the results measured in the graphite sleeve samples previously obtained by INAA & GDMS (Glow Discharge Mass Spectrometry) at CEA Saclay, France, and ICP-MS (CPST, Lithuania) methods. Validation of the ICP-MS method for the nuclear graphite impurity concentration determination was proven. The experimentally obtained RBMK-1500 graphite impurity values in different graphite constructions were compared with other measurements and new limits of the possible maximal concentrations of nuclear RBMK graphite impurity concentrations were obtained.

A stable algorithm for calculating one-group diffusion coefficient in a multi-zone cylindrical cell by the surface pseudo-sources method

In the surface harmonics method, designed to solve the neutron transport equation of in the core of a nuclear reactor, the distribution function is decomposed by trial functions. The first trial function is the absorption and division matrix. It is responsible for the main processes (absorption and fission) that occur inside the nuclear reactor cell. The second trial function is the diffusion coefficient matrix. It is associated with the processes of neutron flow from one cell to another. It is usually calculated in the P2 – and P4 – approximations of the spherical harmonics method and by the surface pseudo-sources method. In close lattices with high absorption of P2 – and P4 – approximations may not be enough, and the surface pseudo-sources method proves to be unstable in the calculation of multi-zone cylindrical cells. This work is dedicated to the development of stable one-group calculation of the diffusion coefficient by the surface pseudo-sources method first, by eliminating the divergent integrals, and secondly, the modification of the algorithm of matrix factorization for calculating one-group neutron transport equation. Developed program DICOF may use in the complexes of WIMS-SH and SVL for stable calculation of the diffusion coefficients matrix for calculation of the VVER and RBMK reactors.

Estimation of the inventory of14C and other key radionuclides in irradiated RBMK-1500 graphite based on limited measurements and full 3D core modeling

ABSTRACTIgnalina NPP contains two Units with RBMK-1500 reactors. After shutdown, several Unit 1 systems and equipment were radiologically characterized and dismantled. The highest volume of reactor structures is attributed to the graphite stack of the reactor core, radiological characterization of which has not yet been performed. The stack can be visualized as a vertical cylinder 8 m high and 14 m diameter, made up of 2488 columns where each column is made up from several graphite blocks. The total mass of the graphite stack blocks is about 1700 tonnes. Therefore, the main goal of work reported in this paper was to estimate the inventory of14C and other key radionuclides in the irradiated graphite by a combination of activity measurements and full 3D reactor graphite stack neutron activation modeling. Obtained results show that, based on the combination of modeling and measurement techniques, the total inventory of14C in graphite stack is estimated at 3.22×1014Bq at 9 years after Unit 1 reactor final shutdown.14C activity is the highest among the analyzed radionuclides; the second highest is60Co (~6 times lower).

Flow-accelerated corrosion rate and residual life time estimation for the components of pipeline systems at nuclear power plants based on control data

As of today, large volumes of data related to non-destructive operational control are accumulated on NPPs. For ensuring safe operation of power units, optimization of scope and scheduling operational control it is necessary to continue development of guidance documents, software products, methodological guidance and operational documentation (Baranenko et al. 1998, Gulina et al. 2013, Recommendation (NSAC-202L-R4) 2013). Approaches are examined to assessment of the rate of erosion-corrosion wear (flow-accelerated corrosion - FAC) according to the data of operational control. The present study was performed based on the data of thickness gauging of different elements of pipelines of NPPs with different types of reactor. Further development of ideas exposed in (Baranenko et al. 2016) allowed revealing specific features of ECW processes on straight sections, bends and in the zones adjacent to weld joints of pipelines of NPPs equipped with VVER and RBMK reactors. Presence of the process of deposition of corrosion products on internal surfaces of pipeline walls results in the fact that residual lifetime of elements nominally increases due to deposition. However, real wall thickness under the layer of deposits is unknown just as the initial wall thickness is unknown as well. Investigation implemented in the present study is aimed at the substantiation of the methodology of calculation of FAC rate according to the data of operational control for the purpose of drawing calculation results closer to the reality keeping conservatism. Uniform approach to the assessment of FAC rate in the examined elements of pipelines was developed. Methodologies for evaluation of correction coefficients taking into account dimensional technological tolerances, special features of geometry of the element, as well as effect of deposits on the results of thickness measurements were suggested based on the data of operational control and industry standards. The implemented studies demonstrated efficiency of the developed procedures for pipeline welding zones. Analysis of known and newly developed procedures was performed for bends and ranking of these procedures according to the criterion of “conservatism of evaluation of residual lifetime” was executed. Introduction of correction coefficients allows enhancing conservatism of calculations of lifetime characteristics as compared with calculations performed on the basis of nominal values of thicknesses; the result depends on the type and dimensions of the element, its geometry, as well as on the type of reactor.

Evaluation of the primary displacement damage in the neutron irradiated RBMK-1500 graphite

Numerical simulations of the neutron induced primary displacement damage in the RBMK-1500 reactor graphite were performed using the Monte Carlo technique based on the Norgett-Robinson-Torrens (NRT) model. The sequence of the displacement evaluation consists of the reactor core modelling using MCNP6 and GEANT4 codes and output parameters, such as recoil ion and neutron flux distribution used by SRIM2013 and NJOY2016 codes, respectively, to evaluate displacements per atom. The comparisons were made between these codes in order to evaluate the accuracy of each approach. The neutron flux in the spent RBMK-1500 graphite was sufficiently correctly represented by both MCNP6 and GEANT4 models. The comparison between SRIM and GEANT4 codes shows that GEANT4 predicts fewer vacancies by more than 16% than SRIM in the ion energy range from 1 keV to 1 MeV. The neutron energy required to produce recoils due to scattering is around 80 eV. Below this energy displacements occur mostly due to neutron capture reactions. The average displacement rate was found to be around 0.51 displacements per atom per full power year, which corresponds to the 1.36·1014 n/cm2s neutron flux.

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.

Discrimination of plutonium from thermal reactors in the frame of nuclear forensics

The potential discrimination of 1st generation Pu, averaged over different assemblies of spent fuels from thermal reactors, has been assessed through a simulated fingerprinting procedure. Plutonium, retrieved from spent fuels of burnup in the range 5–50 GWd/MTU and different 235U enrichments at charge, has been considered. The study was based on the Pu isotopic composition, as the required characteristic signatures, retrieved from the reprocessing of spent fuels from PWR, VVER, BWR, RBMK, AGR, MAGNOX and CANDU reactors, charged with different assemblies. The procedure has resolved the Pu of the spent fuels from the different reactors and burnup range considered. The Pu from the different PWR, VVER and BWR spent fuels were clustered together in accordance with the reactor type and the initial enrichment of the fuels. Regarding the AGR, MAGNOX, RBMK and CANDU reactors, the Pu from the spent fuels are discriminated between them and from the LWR reactors considered.

Исследование газоводной смеси в полости гильзы кластерного регулирующего органа реактора РБМК-1000

Исследование газоводной смеси в полости гильзы кластерного регулирующего органа реактора РБМК-1000

Numerical Study of the Dynamic Strength of RBMK-1000 Stack with Fuel Channel Rupture

The strength of the stack in a RBMK-1000 reactor facility with a rupture of a fuel channel is studied. Local models of unirradiated and irradiated stack, comprising an assembly consisting of 81 columns, and a complete model of the original stack are developed to calculate the parameters of RBMK-1000 core degradation. Numerical studies of the nonstationary stress-strain state of the initial unirradiated stack and the stack degraded by irradiation one year after the repair-restorative work, which confirm the core strength to be adequate, are presented.

Computational and Experimental Validation of the Serviceability of RBMK Core Elements Under Graphite Stack Distortion Conditions

Computational and experimental validation of the serviceability of zirconium channels in the reactor core according to strength criteria was conducted in order to ensure safe operation of the RBMK reactor facility under graphite stack deformation conditions. A complex of computational and experimental studies confirmed that the requirements of the operative federal norms and regulations for objects using nuclear power are met

Numerical analysis of thermal and hydro dynamical processes in upper fuel channel part of boiling water reactor

This article presents results of 3D numerical analysis of the thermal and hydro dynamical processes which take place inside the fuel channel of the RBMK-1500 reactor during interaction between coolant and fuel assembly. Numerical modelling was made for three different operational regimes of reactor: for minimal (0.91 MW), average (2.53 MW) and for the maximal power (3.50 MW) levels. Simulation was made using ANSYS CFX program package in order to evaluate coolant’s pressure, temperature, flow velocity and steam mass fraction changes at the distance grids (spacers) and at the heat transfer intensification grids (intensifiers) along and across the fuel channel. In order to create CFX mesh, a tetrahedral finite element type with 10 million elements was chosen. The k-e turbulence model and homogeneous model of the heat transfer (HEM) were applied for numerical simulation. The obtained results help better to understand the processes which have a direct influence on the reactor safety. Information about the peculiarities of steam generation process inside the fuel channel (temperature change, pressure drop, flow velocity, steam mass fraction) enables to identify the most loaded points of the fuel channel and to make necessary design and operational improvements (location of spacers and intensifiers; core loading and etc.). The results can be used for predicting of hydrogen formation process and for the forecast of the decrease of zirconium alloy strength.DOI: http://dx.doi.org/10.5755/j01.mech.23.5.16457

Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

Моделирование реакторного графита РБМК-1000 с учетом некоторых особенностей микроструктуры

Eleven RBMK-1000 reactors (the RBMK-1000 denotes a large-capacity pressure tube nuclear reactor for a power unit with a 1000 MW electric power output) are presently in operation in the Russian Federation. At the time when these reactors were only commissioned, their service life was assigned to be 30 years. However, the replacement capacities had not been commissioned, and studies of the state of reactor elements revealed that the reactor service life was far from having been exhausted. Therefore, it was decided to extend the operation of these reactors beyond their initially assigned lifecycle. One of the measures used to justify that decision was numerical simulation of the graphite stack behavior, because this element determines the reactor service life in view of impossibility to repair, anneal it, etc. In turn, to achieve the required accuracy of graphite stack numerical analysis, it is necessary to realistically calculate the state of individual graphite blocks from which it is assembled. This is a rather intricate problem because it involves the need to carry out combined thermal, neutronic, and strength calculations. In addition, the analysis must take into account that the graphite used in the RBMK reactor is an anisotropic material. In addition, there was a need to solve a very complicated problem of taking into account the two-phase structure of the graphite, because it consists of a binder and filler. The article presents a study of the effect the scale factor has on the reactor graphite behavior, which is taken into account by means of a phenomenological model. It is assumed that the microstresses arising in a graphite item have a significant influence on its behavior. The calculations were carried out, taking the RBMK reactor graphite block as an example, using a 3D finite element technique and a computer program that takes into account the anisotropy of the reactor graphite behavior under the effect of thermal and radiation factors, its creep, and possible occurrence of cracks.