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Sample records for steel-15kh2mfa

  1. Radiation embrittlement of WWER 440 pressure vessel steel and of some improved steels by western producers

    International Nuclear Information System (INIS)

    Koutsky, J.; Vacek, M.; Stoces, B.; Pav, T.; Otruba, J.; Novosad, P.; Brumovsky, M.

    1982-01-01

    The resistance was studied of Cr-Mo-V type steel 15Kh2MFA to radiation embrittlement at an irradiation temperature of around 288 degC. Studied was the steel used for the manufacture of the pressure vessel of the Paks nuclear reactor in Hungary. The obtained results of radiation embrittlement and hardening of steel 15Kh2MFA were compared with similar values of Mn-Ni-Mo type steels A 533-B and A 508 manufactured by leading western manufacturers within the international research programme coordinated by the IAEA. It was found that the resistance of steel 15Kh2MFA to radiation embrittlement is comparable with steels A 533-B and A 508 by western manufacturers. (author)

  2. Mechanical and fracture properties at impact loading of selected steels for nuclear power engineering

    International Nuclear Information System (INIS)

    Buchar, J.; Bilek, Z.

    1988-01-01

    The possibilities are briefly characterized of experimental research of mechanical and fracture properties of steels used in nuclear power engineering. Attention is paid to plastic deformation and the assessment of fracture formation during impact loading. The results are reported for steels 15Kh2MFA and 10GN2MFA. For steel 15Kh2MFA the effect was also studied of neutron radiation at different temperatures. From the theory developed for non-irradiated material 10GN2MFA, a prediction is made within the original model of the fracture stress value for steel 15Kh2MFA in both non-irradiated and irradiated states. The conclusion is arrived at that the existing methods of assessing steel properties at impact load allow obtaining knowledge of all significant effects during actual stress, this using only small specimens of the materials. (Z.M.). 4 figs., 8 refs

  3. Creep and long-term strength of heat-resistant steels with different structures with the account taken of the type of stress deviator

    International Nuclear Information System (INIS)

    Giginyak, F.F.; Dragunov, Yu.G.; Mozharovskaya, T.N.; Titov, V.F.

    1993-01-01

    The results of the experimental investigations into creep and long-term strength of heat-resistant steels 15Kh2MFA and 15Kh2NMFA in the initial state and after heat-treatment simulating the metal irradiation embrittlement at the end of the product service date under static loading at the complex stress state and at high temperatures are presented. The experimentally substantiated equations of state describing creep and long-term stability of materials taking into account the type of the stress state are derived. (author)

  4. Cyclic crack resistance of anticorrosion cladding-15Kh2MFA steel joint

    International Nuclear Information System (INIS)

    Zvezdin, Yu.I.; Nikiforchin, G.N.; Timofeev, B.T.; Zima, Yu.V.; Andrusiv, B.N.

    1985-01-01

    Cyclie crack resistance of transition zone in austenitic cladding steel 15Kh2MFA joint is studied, taking into account the geometry of fatigue cracks, fracture micromechanism and crack closure effect. Kinetics of crack development from the cladding to the basic metal and vice versa is considered. Microstructure of transition zone is investigated. The results obtained are considered as applied to WWER. It is emphasized, that the braking of fatigue cracks is observed at low asymmetry of loading cycle. Increased loading asymmetry accelerates sharply the alloy fracture due to the growth of subcladding crack, at that, the direction of crack propagation and the structure of transition zone are not of great importance

  5. Study of brittle crack jump rate using acoustic emission method

    International Nuclear Information System (INIS)

    Yasnij, P.V.; Pokrovskij, V.V.; Strizhalo, V.A.; Dobrovol'skij, Yu.V.

    1987-01-01

    A new peocedure is elaborated to detect brittle jumps of small length (0.1...5mm) occuring both inside the specimen and along the crack front under static and cyclic loading using the phenomena of acoustic emission (AE). Recording of the crack start and stop moments with an AE sensor as well as evaluation of the brittle crack jump length by the after-failure specimen fracture make it possible to find the mean crack propagation rate. Experimental dependences are obtained for the crack propagation rate with a brittle crack jump in steel 15Kh2MFA (σ B =1157 MPa, σ 0.2 =100 MPa) at 293 K and under cyclic loading as a function of the jump length and also as a function of the critical stress intensity factor K jc i corresponding to the crack jump

  6. Effect of neutron irradiation on the properties of the repair welds of the 15Kh2MFA steel

    International Nuclear Information System (INIS)

    Morozov, A.M.; Khachaturyants, L.V.

    1986-01-01

    The authors studied the effect of neutron irradiation on the tendency of the metal belonging to the heat affected zone of the weld toward brittle fracture (an increase in the critical temperature of brittleness). For comparison, the authors studied the radiation embrittlement of the original base metal (steel 15Kh2MFA) subjected to the conventional heat treatment of the reactor frames consisting of hardening and high-temperature tempering. Along with these materials, the radiational embrittlement of the base metal in the rehardened condition without tempering was studied. It was concluded that the presence of the regions repaired according to this technology and located in the frame at the level of the reactor core does not pose the problem of decreased resistance to brittle fracture

  7. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kocik, J.; Keilova, E.

    1993-01-01

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs

  8. Study of susceptibility to hydrogen embrittlement of welded joints of large WWER reactor vessels at different temperatures

    International Nuclear Information System (INIS)

    Mazel', R.E.; Kuznetsova, T.P.; Grinenko, V.G.; Sapronova, M.N.

    1977-01-01

    The effect is studied of hydrogen and a coolant of WWER on the susceptibility to brittle fracture of welded joints from steels 15Kh2MFA and 15Kh2NMFA obtained by automatic submerged arc welding with the use of the welding materials of different purity. The effect of hydrogen (concentration range 0.5-7.5 cm 3 /100 g, testing temperatures 20, 70 and 325 deg C) and the coolant (pressures up to 120 atm, temperatures 20-350 deg C) have been estimated by the fracture work during static bending tests. It is shown that the purification of the welding materials enhances the fracture properties by about a factor of 2. Hydrogenation results in a sharp drop (by about a factor of 3) of the fracture work. The increased testing temperature (up to 325 deg C) is accompanied by disappearance of the effect of hydrogen embrittlement, which is explained by an increase in the diffusion mobility of atomic hydrogen. Under the action of the coolant the fracture work shows a two-fold decrease, while the pressure being increased up to 100 atm leads to greater fracture work decrease

  9. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Kocik, J; Keilova, E [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding) to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs.