RBMK Reactors Research Articles

Modeling of the radiation doses during dismantling of RBMK-1500 reactor emergency core cooling system large diameter pipes

Personnel radiation safety is one of the most important issues during the dismantling of nuclear installations. In this paper, results of modeling radiation doses during the dismantling of the large diameter pipes from the emergency core cooling system of RBMK-1500 reactor at Ignalina NPP are presented. The effective doses to the workers are modeled for four dismantling alternatives in order to propose the optimal dismantling approach. The impacts of the cutting technology and individual respiration protection on effective doses are analyzed. The total effective personnel doses are obtained by summing the effective personnel doses from various sources of exposure, i.e., direct radiation from radioactive equipment, internal radiation due to inhalation and ingestion of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting. The collective effective doses and their uncertainties are analyzed using VISIPLAN and MATLAB codes.

Improvement of Algorithms for Pressure Maintenance Systems in Drum-Separators of RBMK-1000 Reactors

The main tasks and challenges for pressure regulation in the drum-separators of RBMK-1000 reactors are described. New approaches to constructing algorithms for pressure control in drum-separators by electro-hydraulic turbine control systems are discussed. Results are provided from tests of the operation of modernized pressure regulators during fast transients with reductions in reactor power.

Upgrading the SPP-500-1 moisture separators-steam reheaters used in the Leningrad NPP turbine units

The specific features of existing designs of moisture separators-steam reheaters (MSRs) and experience gained with using them at nuclear power plants are considered. Main factors causing damage to and failures of MSRs are described: nonuniform distribution of wet steam flow among the separation modules, breakthrough of moisture through the separator (and sometimes also through the steam reheater), which may lead to the occurrence of additional thermal stresses and, hence, to thermal-fatigue damage to or stress corrosion cracking of metal. MSR failure results in a less efficient operation of the turbine unit as a whole and have an adverse effect on the reliability of the low-pressure cylinder’s last-stage blades. By the time the design service life of the SPP-500-1 MSRs had been exhausted in power units equipped with RBMK-1000 reactors, the number of damages inflicted to both the separation part and to the pipework and heating surface tubes was so large, that a considerable drop of MSR effectiveness and turbine unit efficiency as a whole occurred. The design of the upgraded separation part used in the SPP-500-1 MSR at the Leningrad NPP is described and its effectiveness is shown, which was confirmed by tests. First, efforts taken to achieve more uniform distribution of moisture content over the perimeter and height of steam space downstream of the separation modules and to bring it to values close to the design ones were met with success. Second, no noticeable effect of the individual specific features of separation modules on the moisture content was revealed. Recommendations on elaborating advanced designs of moisture separators-steam reheaters are given: an MSR arrangement in which the separator is placed under or on the side from the steam reheater; axial admission of wet steam for ensuring its uniform distribution among the separation modules; inlet chambers with an extended preliminary separation system and devices for uniformly distributing steam flows in the separator; separated layout of the of the separator and steam reheater; and use of transversely finned tube bundles for organizing cross flow of steam over the tubes.

Results of an Analysis of Field Studies of the Intrinsic Dynamic Characteristics Important for the Safety of Nuclear Power Plant Equipment

A classification of the equipment important for the safety of nuclear power plants is proposed in terms of its dynamic behavior under seismic loading. An extended bank of data from dynamic tests over the entire range of thermal and mechanical equipment in generating units with VVER-1000 and RBMK-1000 reactors is analyzed. Results are presented from a study of the statistical behavior of the distribution of vibrational frequencies and damping decrements with the “small perturbation” factor that affects the measured damping decrements taken into account. A need to adjust the regulatory specifications for choosing the values of the damping decrements with specified inertial loads on equipment owing to seismic effects during design calculations is identified. Minimum values of the decrements are determined and proposed for all types of equipment as functions of the directions and natural vibration frequencies of the dynamic interactions to be adopted as conservative standard values in the absence of actual experimental data in the course of design studies of seismic resistance.

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.

Features of the application of the Monte Carlo method to calculations for large RBMK reactors and to model correction on the basis of data from in-core detectors

Statistical errors in sampling neutron fields in physically large systems like an RBMK are analyzed both qualitatively and quantitatively. Recommendations concerning the choice of parameters for calculations are given. A new procedure for Monte Carlo RBMK calculations with model corrections on the basis of data from in-core detectors is proposed. Dedicated software based on the CUDA software and hardware platform is developed for computational research. Results of testing the procedure and software in question via calculations for real RBMK reactors are discussed.

High-Temperature Heat-Treatment of the Welded Joints of Du300 Austenitic Pipelines at 900°C

The high-temperature heat treatment of the welded joints of Du300 austenitic pipelines on straight sections and lead-aways, welded joints of Du300 pipelines to pipe connections for equipment and fittings and welded joints in the zones of construction walkways is described. It was implemented in the welded joints in a real Du300 pipeline and full-scale mockup of a Du300 pipeline and the pipe connections of the intake manifold in an operating power-generation unit of the Leningradskaya NPP with an RBMK-1000 reactor, which was shut down for midlife repairs. The technology suppresses the sensitization of the metal in the near-joint zone to below the proneness toward intercrystallite corrosion cracking under stress (coefficient of potentiodynamic reactivation below 1%) and decreases the residual stresses by 70–80%.

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.

Influence of Soaking Temperature and Time on Microstructure and Mechanical Properties of Water Quenched Zr–2.5Nb Alloy

In heavy water cooled nuclear reactors, Zr–2.5Nb alloy is used for pressure tubes under cold worked and stress relieved (CWSR) condition in Canadian Deuterium Uranium (CANDU) reactors and under quenched and aged condition in RBMK and Fugen reactors. Because of its ease of use and the lower cost of fabrication, most of the work reported in literature is focused on the development and characterization of cold worked and stress relieved pressure tube material. However, recent work showed that the tubes manufactured using quenched and aged route showed a lower and predictable rate of in-reactor deformation during reactor operation as compared to those fabricated using the CWSR route. One of the important stages in the fabrication of heat treated pressure tube material is solution heat treatment (SHT), which governs the microstructure and hence, the mechanical properties of pressure tubes. Usually, SHT is carried out in two phase (α + β) region at a temperature closer to the β transus. The present work characterizes the influence of different SHT parameters such as soaking temperature (850, 870, and 890°C) and duration (15 and 30 min) on microstructure, texture, and Nb partitioning between phases and mechanical properties such as tensile properties, hardness, and fracture toughness. Fracture toughness parameters were determined as per ASTM standard E1820-11. Optical microscopy was used for microstructure, X-ray diffraction for texture, and EPMA for Nb partitioning investigation. The fracture surfaces were examined using SEM.

Use of software tools for calculating flow accelerated corrosion of nuclear power plant equipment and pipelines

The results obtained from calculations of flow accelerated corrosion of equipment and pipelines operating at nuclear power plants constructed on the basis of PWR, VVER, and RBMK reactors carried out using the EKI-02 and EKI-03 software tools are presented. It is shown that the calculation error does not exceed its value indicated in the qualification certificates for these software tools. It is pointed out that calculations aimed at predicting the service life of pipelines and efficient surveillance of flow accelerated corrosion wear are hardly possible without using the above-mentioned software tools.

Comparison of Radionuclide Releases from a Conceptual Geological Repository for RBMK-1500 and BWR Spent Nuclear Fuel

Approximately 22 600 spent nuclear fuel (SNF) assemblies originating from the RBMK-1500 reactor of the Ignalina nuclear power plant in Lithuania need to be managed and disposed of safely. Generic investigations of RBMK-1500 SNF disposal options in Lithuania were initiated. This paper presents insights on RBMK-1500 SNF disposal in crystalline rocks gained during participation in the International Atomic Energy Agency Coordinated Research Project “The Use of Numerical Models in Support of Site Characterization and Performance Assessment Studies for Geological Repositories,” as well as in the Lithuanian Science Development Program. The research was focused on the analysis of disposal behavior of different SNF types under generic geological conditions and for a one-canister defect scenario with two different corrosion rates. A comparison of peak fluxes from the near field for Lithuanian RBMK-1500 and Swedish boiling water reactor SNF revealed differences that are not directly proportional to the differences in SNF inventory.

Analysis of alternatives for radwaste management during dismantling of the equipment in building V1 at Ignalina NPP

Ignalina NPP was operating two RBMK-1500 reactors. Unit 1 was closed at the end of 2004, and Unit 2 was closed at the end of 2009. Now preparation for dismantling activities, including building V1, is ongoing. Building V1 is located on the west side of Unit 1. It is a nine-level building with more than 200 rooms. Part of Unit 1 auxiliary systems are located in Building V1. Most systems in this building are contaminated. After situation analysis in Building V1 at Ignalina NPP and collection of the primary information related to the components’ physical and radiological characteristics, location and other data, two alternatives for radwaste management during dismantling were formulated and evaluated: the first, when the decontamination of dismantled components is performed (where it is reasonable) and the second one, when no decontamination of dismantled components is performed and after dismantling the components are routed to appropriate waste storage or disposal sites. To select the preferable alternative, MCDA method – AHP (Analytic Hierarchy Process) was applied. Hierarchical list of decision criteria, necessary for assessment of alternatives performance, were formulated. Quantitative decision criteria for these alternatives were analyzed using software DECRAD, which was developed at Lithuanian Energy Institute Nuclear engineering laboratory. Qualitative decision criteria were evaluated using expert judgment. The results have shown that the first alternative (decontamination is performed) is better than the second one (no decontamination is performed).

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.

Prediction of radiation doses during the dismantling of a maintenance cooling reservoir of RBMK-1500 reactor

*Decommissioning of the Ignalina Nuclear Power Plant involves multiple problems. Personnel radiation safety during the dismantling activities is one of them. In order to assess the optimal personnel radiation safety, the modelling is performed by means of computer code “VISIPLAN 3D ALARA Planning tool” developed by SCK CEN (Belgium). In this paper it was used for the evaluation of direct gamma radiation from the radioactive equipment to personnel when carrying out dismantling activities and selection of gamma radiation shielding geometry (structure), shielding materials and positions. A modelling of radiation doses during the dismantling of the maintenance cooling reservoir (MCR) of RBMK-1500 reactor has been performed in this paper. The basic MCR model’s development was based on internal contamination and the existing geometrical data. The assessment of workers exposure was performed to comply with ALARA. The effective doses to the workers were calculated for different strategies of reservoir dismantling. The impacts of dismantling tools, shielding types, variation of parameters such as shielding distance, shielding installation time and exhaust ventilation flow rate during the dismantling of reservoir on effective doses were analyzed. The total effective personnel doses were obtained by summarizing the effective personnel doses from various sources of exposure, i.e., direct radiation from radioactive equipment, internal radiation due to inhalation of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting in premises.

Implementation of the indirect assessment method for 129I and 135Cs accumulation in the RBMK-1500 reactor coolant purification system

The transport of long-lived radionuclides 129I and 135Cs in technologic chains of the Ignalina NPP and their accumulation in by-pass purification system filtering materials was investigated by spectrometric measurements and mathematical modelling. The database of regular gamma spectrometric measurements performed at the NPP was analysed. Results of gamma spectrometric measurement of specific activities of 131I and other short-lived isotopes of iodine as well as of 134Cs and 137Cs isotopes have been analysed and parameters specific to the reactor operation have been estimated. Accumulated activities of 129I and 135Cs in spent filtering materials of the reactor by-pass purification system as well as specific activities are evaluated. Specific activities are compared with preliminary acceptance criteria for a near surface repository which is at the stage of design at the Ignalina NPP. Moreover, scaling factors for these radionuclides for the by-pass purification system radioactive waste are assessed.

Analysis of radwaste management alternatives during dismantling of Ignalina NPP systems with low level contamination

Abstract Ignalina NPP was operating two RBMK-1500 reactors which are under decommissioning now. In this paper, analysis on radwaste management alternatives during the dismantling of systems with low level contamination and different types of components in buildings 117/1 and V1 are presented. After situation analysis and collection of the primary information related to components' physical and radiological characteristics, location and other data, two alternatives for radwaste management during the dismantling were formulated and evaluated: the first one (A1) when the decontamination of the dismantled components is performed (if it is reasonable), and the second one (A2) when no decontamination of the dismantled components is performed and after the dismantling, the components are routed to appropriate waste storage or disposal sites. To select the preferable alternative, MCDA method – AHP (Analytic Hierarchy Process) is applied. Hierarchical lists of decision criteria, necessary for assessment of alternatives performance, are formulated. Quantitative decision criteria values for these alternatives are calculated using software DECRAD, which was developed by Lithuanian Energy Institute Nuclear Engineering Laboratory. Qualitative decision criteria are evaluated using expert judgment. Analysis results show that alternative A1 has a preference against alternative A2.

Use of radiation monitors for personnel contamination detection

It is shown that the TSRM61 and TSRM82 radiation monitors produced at the Dukhov All-Russia Research Institute of Automatics (VNIIA) can be used for detecting the limit of the contamination of the clothing, footwear and skin of the personnel in radiation enterprises. The main radionuclides corresponding to the decay products of VVER, RBMK, and fast reactor fuel are examined.

An investigation of the effect of hydrodynamics of flow on erosion-corrosion resistance of components of a feedwater control assembly used in power units of nuclear power plants equipped with RBMK-1000 reactors

The results of computational hydrodynamic simulation of the flow of a working medium in the flow path of the feedwater control assembly used in power units of nuclear power plants (NPPs) equipped with the reactors of the RBMK-1000 type are presented It was established that the rate of control valve opening has an influence on the position of the areas of the intense local erosion-corrosion thinning of metal of the diffuser section downstream of the valve.

Preventing inadmissible erosion-corrosion thinning from occurring in the diffuser segments of feedwater supply control systems of power units at nuclear power stations equipped with RBMK-1000 reactors

Results obtained from investigations of erosion-corrosion processes that occur during operation of the feedwater supply control systems used in power units of nuclear power stations equipped with RBMK-1000 reactors and the sensitivity of these processes to variations in the chemical composition of metal and in the flow path geometry are presented. It is found that local erosion-corrosion thinning of the walls in the diffuser segments of feedwater supply control systems occur mainly due to intense mass transfer in the near-wall region taken in combination with a low content of chromium. Hydrodynamic simulation was carried out, and it was shown based on its results that local erosion-corrosion thinning of the walls of pipeline segments downstream of the valves controlling the supply of feedwater to power units of nuclear power stations equipped with RBMK-1000 reactors can be prevented by subjecting them to appropriate modernization. It is found that the above mentioned diffuser parts can be made more resistant to erosion-corrosion wear by keeping the content of chromium in the main metal and weld joints at a level of no less than 0.25% and concurrently reducing the hydrodynamic effect in the zones of weld connections.