Safety assurance of spent nuclear fuel dry storage requires reliable prediction of mechanical properties of fuel element cladding depending on structural changes at all stages of nuclear fuel handling. In this work investigations of hydrides structure in irradiated fuel element cladding made of E110 alloy have been conducted in irradiated state as well as after tests simulating dry storage conditions. Using the program code, the coefficients that best correlate with mechanical properties have been found.

One of the factors capable of resulting in deconfinement of fuel element cladding upon long term storage of spent nuclear fuel (SNF) is reorientation of hydrides. This work estimates the extent of reorientation of hydrates in unirradiated gas filled specimens of fuel element claddings of VVER-1000 reactor (Zr‒1% Nb) under the conditions simulating conditions of dry storage. The specimens with the hydrogen content of 100 and 200 ppm were exposed to thermomechanical tests according to various scenarios at hoop stresses in cladding up to 100 MPa. It has been established that the extent of reorientation of hydrides and related decrease in plasticity of cladding material are insignificant. The most significant factor influencing the extent of reorientation is the level of hoop stresses.

The penetration of atomic hydrogen into the cladding material of fuel elements of WWER-1000 reactors due to interaction with the coolant during operation can subsequently significantly reduce their plasticity characteristics with a decrease in temperature during long-term dry storage of spent nuclear fuel (SNF) due to the formation of brittle hydrides. The morphology of hydrides influenced by the hydrogen content, storage temperature and circumferential stresses plays a decisive role in the embrittlement of the fuel cladding material. Associated radial hydrides are of particular danger; they constitute the most favorable path for crack propagation.In the present work, thermomechanical tests of irradiated fuel claddings samples made of the E110 alloy were carried out, simulating normal and emergency conditions of long-term dry storage. It was shown that under the conditions considered, the formation of a significant amount of radially oriented hydrides was observed, leading to degradation of mechanical properties (embrittlement) fuel cladding.

In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC "Kurchatov Institute" has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to-brittle transition temperature shift is given.

In this paper comprehensive studies of structure and properties of VVER-1000 RPV steels after the accelerated irradiation to fluences corresponding to extended lifetime up to 60years or more as well as comparative studies of materials irradiated with different fluxes were carried out. The significant flux effect is confirmed for the weld metal (nickel concentration⩾1.35%) which is mainly due to development of reversible temper brittleness. The rate of radiation embrittlement of VVER-1000 RPV steels under operation up to 60years and more (based on the results of accelerated irradiation considering flux effect for weld metal) is expected not to differ significantly from the observed rate under irradiation within surveillance specimens.

In a temperature range of 280–320°C, the mechanism and kinetics of segregation of impurities in steels have yet remained insufficiently studied. Under these conditions diffusion of impurities in the bulk of steel grains practically ceases, and for describing the kinetics of the process it is incorrect to use the Langmuir-McLean equation. In this work we put forward two new approaches to describe the mechanism and kinetics of phosphorus segregation in steels: a model of sequential changes in the state of phosphorus based on first-order reactions and a model of diffusion redistribution of phosphorus between boundaries of carbide precipitates, structure defects, and boundaries of steel grains. A comparative analysis of the suggested models has been conducted, and estimates of the kinetics of segregation based on them have been made; these estimates have been compared with the experimental results obtained in the temperature range of 280–320°C for test times to ∼20 years. It has been shown that these models fairly well describe the experimental kinetics of phosphorus segregation in boundaries of steel grains.

In this paper the influence of fast neutron flux on the structural features and properties of VVER-1000 reactor pressure vessel steels was studied. It is shown that for high Ni steels the flux effect is due to hardening and non-hardening mechanisms of radiation embrittlement.

Service life of VVER-type nuclear reactor is limited by decrease in brittle fracture resistance of reactor pressure vessel produced of low-alloy low-carbon steel under effect of irradiation and/or elevated temperatures. In this work fracture surfaces were studied by Auger-electron spectroscopy in order to estimate the contribution of intergranular embrittlement to the degradation of reactor pressure vessel steels under the influence of operating conditions. It was demonstrated that irradiation induced segregation leads to an increase of P content in grain boundaries that promotes intergranular brittle fracture on fracture surfaces. The similar effect but to a lesser degree was shown in the case of long-term temperature exposure. The grain boundary structure was examined and an effect of carbides located on the grain boundaries is supposed due to increased phosphorus segregation on carbide/matrix interface boundaries.

In this paper the influence of structural parameters on the tendency of steels to reversible temper embrittlement was studied for assessment of performance properties of reactor pressure vessel steels with extended service life. It is shown that the growth of prior austenite grain size leads to an increase of the critical embrittlement temperature in the initial state. An embrittlement heat treatment at the temperature of maximum manifestation of temper embrittlement (480 °C) shifts critical embrittlement temperature to higher values due to the increase of the phosphorus concentration on grain boundaries. There is a correlation between phosphorus concentration on boundaries of primary austenite grains and the share of brittle intergranular fracture (that, in turn, depends on impact test temperature) in the fracture surfaces of the tested Charpy specimens.

After ten years of operation, the absorption tower at a nitric acid plant was failed due to stress corrosion cracking. SCC was caused by combination of both welding residual tensile stresses and chloride ions accumulated due to rain and/or vapor contaminated with chloride. Irregularity of fillet weld surface helped in occurrence of cycles of wetting and drying conditions of weld surface that in turn resulted in chloride concentration. The problem has been solved by changing design of the fixation system of the rain water collector in order to avoid combination of tensile stresses and chloride, the condition necessary for occurrence of SCC. In other words, fixation of rain water’ collector to the tower’ shell was done using mechanical means instead of fillet welds that in turn result in eliminating tensile stresses and reducing chloride concentration.

IRRADIATION STIMULATED INTERGRANULAR SEGREGATION IN VVER-1000 REACTOR PRESSURE VESSEL MATERIALS O.O. Zabusov, M.A. Saltykov, B.A. Gurovich, E.A. Kuleshova, S.V. Fedotova, D.A. Zhurko Auger-electron spectroscopy (AES) experimental data on changes of grain boundary segregation changes in VVER-1000 reactor pressure vessels (RPV) surveillance specimens both weld and base metals in initial and irradiated states and after continuous thermal exposure are presented. The problem of the reversible temper embrittlement effect at recovery annealing proceedure of these RPV is considered. The analysis of AES results together with fractographic study and mechanical tests results made it possible to work out an approach to assessment of intergranular embrittlement contribution in changes of mechanical properties of different RPV elements by the end of extended service life.

MICROSTRUCTURE PECULARITIES OF VVER-1000 REACTOR PRESSURE VESSEL MATERIALS FRACTURE SURFACES M.A. Saltykov, O.O. Zabusov, B.A. Gurovich, M.A. Artamonov, A.P. Dementjev, E.A. Kuleshova, S.V. Fedotova, D.A. Zhurko The experimental data are presented on determination of phosphorus segregation on grain boundaries and precipitation surfaces located on grain boundaries of VVER-1000 reactor pressure vessel (RPV) materials subjected to continuous thermal exposures using Auger-electron spectroscopy. The influence of precipitates on the brittle transgranular fracture of an irradiated VVER-1000 RPV weld metal is shown. Auger peaks fine structure of carbon on grain boundaries was carried out, a qualitative evaluation of carbon role in the brittle intergranular fracture mechanism is presented.

The results of VVER-440 steel Sv-10KhMFT and VVER-1000 steel SV-10KhGNMAA investigations by transmission electron microscopy, scanning electron microscopy, Auger-electron spectroscopy and mechanical tests are presented in this paper. The both types of weld metals with different content of impurities and alloying elements were studied after irradiations to fast neutron (E>0.5MeV) fluences in the wide range below and beyond the design values, after recovery annealing procedures and after re-irradiation following the annealing. The distinctive features of embrittlement kinetics of VVER-440 and VVER-1000 RPV weld metals conditioned by their chemical composition differences were investigated. It is shown that the main contribution into radiation strengthening within the design fluence can be attributed to radiation-induced precipitates, on reaching the design or beyond design values of fast neutron fluencies the main contribution into VVER-440 welds strengthening is made by radiation-induced dislocation loops, and in case of VVER-1000 welds – radiation-induced precipitates and grain-boundary phosphorous segregations. Recovery annealing of VVER-440 welds at 475°C during 100h causes irradiation-induced defects disappearance, transformation of copper enriched precipitates into bigger copper-rich precipitates with lower number density and leads to almost full recovery of mechanical properties followed by comparatively slow re-embrittlement rate. The recovery annealing temperature of VVER-1000 welds was higher – 565°C during 100h – to avoid temper brittleness. The annealing of VVER-1000 welds leads to almost full recovery of mechanical properties due to irradiation-induced defects disappearance and decrease in precipitates number density and grain-boundary segregation of phosphorus. The re-embrittlement rate of VVER-1000 weld during subsequent re-irradiation is at least not higher than the initial rate.