Evaluation of pipe whip impacts on neighboring piping and walls of the Ignalina Nuclear Power Plant

Stress corrosion cracks have been discovered in Group Distribution Headers (GDH) at the Ignalina and Chernobyl Nuclear Power Plants. This increases the probability that a guillotine pipe break can occur that creates a whipping pipe (GDH) with the potential to damage surrounding structures—i.e. adjacent GDH and its attached piping or adjacent reinforced concrete compartment wall. The GDH is the most important component for reactor safety in case of an accident. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the ECSS into the GDH.Presented in this paper is the transient analysis of a Group Distribution Header following a guillotine break at the blind end of the header. Using a very conservative force loading function, the transient response of a whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is obtained using finite element methodology.The results of the study, assuming that the impacted GDH does not suffer stress corrosion cracking, indicate that the structural integrity of the compartment should be maintained and failure should not propagate from GDH to GDH.     

Reliability analysis of pipe whip impacts

A probability-based approach is presented as the integration of probabilistic methods and deterministic modelling based on the finite element method. An existing finite element software package was linked to an existing probabilistic package to analyse the complex mechanics that occur during the transient non-linear analysis of impact problems. This methodology is applied to a pipe whip analysis of a group-distribution-header, which results from a guillotine break, and subsequent impact with the adjacent building wall; this is a postulated accident for the Ignalina Nuclear Power Plant RBMK-1500 reactors. The uncertainties of material properties, component geometry data and loads were taken into consideration. The probabilities of failure of the impacted header and of the header support-wall were estimated given uncertainties in material properties, geometrical parameters and loading. The software ProFES was used for the probabilistic analysis and the finite element software NEPTUNE for deterministic structural integrity evaluation. The Monte Carlo Simulation, First Order Reliability method and Response Surface method were used in the probabilistic analysis.     

Structural response analysis of the confinement system of the Ignalina nuclear power plant

Presented is a study of the accident confinement system of the Ignalina nuclear power plant (NPP) with Russian acronym for ‘channelled large power reactor’ (RBMK)-1500 reactors for the ultimate design basis accident. The computation techniques described pertain to the strength analysis of the accident confinement system (ACS): the design of the ACS compartments, materials, and manner of reinforcement. In strength predictions of reinforced concrete to evaluate the carrying-ability of the structure and its performance by the technique of limit analysis. The ACS analysis with finite elements (FE) of several types yielded instantaneous reserve strength factors for static loads. Numerical results of dynamic effects to the strength evaluation for compartment which houses the downcomers of drum separators are presented. The behaviour of pressure and temperature were taken from the results of thermal hydraulics models. Because of the relatively slow rate of loading, the investigation has confirmed that dynamic effects do not contribute additional insight to regular static analysis.     

Geotechnical and hydrochemical properties of sewage sludge

The geotechnical and geochemical properties of sewage sludge from Kaunas city, Lithuania, were investigated to assess whether it could form a low permeability landfill cover. New sludge, sludge stored or kept in the atmosphere and old sludge were studied. It was found that over time the strength of the sewage sludge increased due to self-consolidation and destruction of organic material. The concentration of metals in the sewage extract was considerably lower than in the landfill leachate. It was concluded that if kept in an open polygon for 1–2 years, the sewage sludge can be used as a low permeability landfill cover.     

Prediction of fuel channel—graphite gas-gap behaviour in RBMK reactors

To maintain thermal contact between the fuel assembly and the graphite moderator, RBMK design reactors employ graphite split rings, which are alternatively tight on the pressure tube or tight on the graphite brick central bore. The split in the graphite rings allows a helium/nitrogen gas mixture to flow up the fuel channel. This prevents oxidation of the graphite and can be sampled to detect pressure tube leaks. The initial clearance between the rings and pressure tube or graphite brick is approximately 2.7 mm (1.35 mm each side). Due to material property changes of the pressure tubes and graphite during operation of the reactor, the size of the clearance between the rings and the pressure tube/brick, called the “gas-gap”, varies. Closure of these gaps has been identified as a possible safety case issue by reactor designers and by independent reviews carried out as part of TACIS reviews and as part of the Ignalina Safety Analysis Report. The reasons for this are that gas-gap closure would cause the pressure tube to be tightly gripped by the graphite bricks via the split rings, which could lead to:

• Extra loading on the upper pressure tube zirconium/steel transition joint, particularly during shut down and emergency transients.

• Splitting of the graphite brick, leading to loss of thermal contact between the pressure tube and graphite. As approximately 5.6% of the heat in graphite-moderated reactor is generated within the moderator through neutron and gamma-heating, loss of thermal contact would result in higher graphite temperatures, accelerating the rate of graphite expansion and hence increasing the loading of the core radial restraint.

• Graphite debris may become lodged in inter-brick gaps, leading to increased axial pressure tube loading during shut down and emergency transients.

The authors have carried out deterministic assessments based on the Ignalina RBMK-1500 reactors in Lithuania, modelling the behaviour of the graphite under irradiation and have predicted graphite bore diameter changes that are in good agreement with the measurements of graphite bore diameters taken at Ignalina Nuclear Power Plant (NPP). A probabilistic model has been developed using the actual results of the deterministic calculations with non-linear graphite behaviour. Statistical analysis of the measurements of tube and graphite diameters taken from Units 1 and 2 at Ignalina NPP has been carried out. Further work has been carried out to try to determine the uncertainty inherent in the predictions of the gas-gap closure from the calculations. The overall objective of the studies is to aid prediction of the gas-gap closure process, and help to identify a suitable monitoring strategy for gas-gap closure that could be used for any RBMK reactor.     

Strength evaluation of a steam distribution device in the Ignalina NPP accident localisation system

The Ignalina NPP has a pressure suppression type of confinement, which is referred to as the accident localization system (ALS). The ALS prevents the release of the radioactive material from the NPP to the environment during a loss-of-coolant accident (LOCA). Ten water pools are located in the two ALS towers (five pools in each tower), which separate the dry well from the wet well. These water pools condense the accident-generated steam and prevent high overpressures in the compartments.The steam distribution device (SDD), with the vertical vent pipes (nozzles) that are inserted under the water of the condensing pools, connects the dry well and the wet well. In case of an accident, these components must be capable of withstanding the dynamic loads generated by a LOCA for successful pressure suppression function.This paper presents the transient analysis of the SDD and their connections to the vertical steam corridors following a LOCA. A thermo-hydraulic analysis of the SDD was performed using the state-of-the-art COCOSYS code to determine pressure and temperature histories resulting from a LOCA. The finite element code NEPTUNE was used to evaluate the structural integrity of the SDD and its supporting reinforced concrete wall. Results show that, although portions of the SDD undergo plastic response and the outside surface of the vertical steam corridor reinforced concrete wall cracks, the structural integrity of the SDD and wall are maintained during a LOCA.     

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

This article deals with the investigation of the hydrogen concentration and temperature influence onto mechanical and fracture mechanics characteristics of RBMK-1500 Ignalina NPP unit 2 reactor fuel channel material—Zr–2.5Nb zirconium alloy (TMT-2) at temperatures from ambient up to 300 °C. The investigation of mechanical characteristics was performed on tensile specimens, fracture mechanics characteristics K_Q, , J_IC—on compact specimens (B = 4 mm) of hydrogen-free and saturated by hydrogen (52, 100 and 140 ppm) at 20, 170, 200 and 300 °C. The investigation showed that temperature increasing calls mechanical strength decreasing, whereas the reductions of area increase. Stronger influence of hydrogen concentration onto mechanical characteristics is noticed only at 20–170 °C temperature, however this influence diminishes as the temperature increases and weakest hydrogen influence is given at 300 °C. Fracture toughness characteristics K_Q, more depends on temperature than on hydrogen concentration. Critical J_IC integral values for the specimens containing hydrogen were given lowest at 20 °C, increases when temperature were raised up to 140 °C and were given highest when it reaches 300 °C.The analysis of and J_IC dependence due to the mechanical characteristics of zirconium alloy has showed that the modified plasticity Z_mod = (R_p0.2/R_m)Z satisfactorily approximates the influence of temperature and hydrogen concentration on variation of these characteristics.     

Identification of a sex pheromone produced by female velvetbean caterpillar moth

A sex pheromone produced by female velvetbean caterpillar moths,Anticarsia gemmatalis Hübner, that attracts conspecific males was isolated and identified as a blend of (Z,Z,Z)-3,6,9-eicosatriene and (Z,Z,Z)-3,6,9-heneicosatriene in a ratio of ca. 53, respectively, when combined. The synthesized compounds elicited responses by velvetbean caterpillar moth males equivalent to those elicited by females in both laboratory wind tunnel bioassays and field trapping experiments.Lepidoptera: Noctuidae.Mention of a commercial or proprietary product does not constitute an endorsement by the USDA.Employed through a cooperative agreement between the Department of Entomology & Nematology, University of Florida, and the Insect Attractants, Behavior, and Basic Biology Research Laboratory, Gainesville, Florida 32604.     

Fracture toughness of welded joints materials for main pipelines at Ignalina NPP

This paper deals with an investigation of mechanical and fracture toughness characteristics of welded joint materials used in Ignalina Nuclear Power Plant (NPP) reactor main circulating circuit (MCC) and steam pipelines. Basic metal of MCC group distributing header (GDH) steel 08Ch18N10T (Du-300), its weld metal welded by manual and automatic arc method using the wire SV-04Ch19N11M3 and electrodes EA-100/10U or EA-100/10T, this joint heat-affected zone metal and base metal of the main steam system—steel 16GS (DU-630) and its weld metal welded by manual arc method using the wire SV-08GS2 and electrodes UONI-13/55 were tested.Mechanical properties of welded joints materials—proportional limit (σ_pl), yield (σ_y) and ultimate (σ_u) strength, fracture stress (σ_f) and ductility (Z) (percent reduction of area) of the specimens were determined. Investigation of relative critical stress intensity factor for fixed thickness of the specimen and critical J-integral, J_IC, was performed.The probabilistic investigation of influence of the mechanical properties (σ_pl, σ_y, σ_u) onto fracture toughness characteristics and J_IC for tested materials by using linear regression model with three independent variables was performed.Research enabled to conclude that proposed multivariable regression model with 80% probability (confidence coefficient α = 0.05) has explained reasonably well the dependence of with σ_pl, σ_y, σ_u and it has shown the non-acceptability of probabilistic evaluation of the model with respect to J_IC.     

Analysis of potential waterhammer at the Ignalina NPP using thermal–hydraulic and structural analysis codes

The RBMK (Russian acronym for ‘channeled large power reactor’)-1500 reactors at the Ignalina nuclear power plant (NPP) have a series of check valves in the main circulation circuit (MCC) that serve the coolant distribution in the fuel channels. In the case of a hypothetical guillotine break of pipelines upstream of the group distribution headers (GDH), the check valves and adjusted piping integrity is a key issue for the reactor safety during the rapid closure of check valve. An analysis of the waterhammer effect (i.e. the pressure pulse generated by the valves slamming closed) is needed. The thermal–hydraulic and structural analysis of waterhammer effects following the guillotine break of pipelines at the Ignalina NPP with RBMK-1500 reactors was conducted by employing the RELAP5 and PipePlus codes. Results of the analysis demonstrated that the maximum values of the pressure pulses generated by the check valve closure following the hypothetical accidents remain far below the value of pressure of the hydraulic tests, which are performed at the NPP and the risk of failure of the check valves or associated pipelines is low. Sensitivity analysis of pressure pulse dependencies on calculation time step and check valve closure time was performed. Results of RELAP5 calculations are benchmarked against waterhammer transient data obtained by employing structural mechanics code BOS fluids.