Irradiation Embrittlement Of Steels Research Articles

Self-Recovering Section of RPV Steel Radiation Embrittlement Kinetics as Indication of Material Smart Behavior

Influence of neutron irradiation on reactor pressure vessel (RPV) steel degradation are examined with reference to the possible reasons of the substantial experimental data scatter and furthermore -nonstandard (non-monotonous) and oscillatory embrittlement behavior. In our glance this phenomenon may be explained by presence of the wavelike recovering component in the embrittlement kinetics. We suppose that the main factor affecting steel anomalous embrittlement is fast neutron intensity (dose rate or flux), flux effect manifestation depends on state-of-the-art fluence level. At low fluencies radiation degradation has to exceed normative value, then approaches to normative meaning and finally became sub normative. Data on radiation damage change including through the ex-service RPVs taking into account chemical factor, fast neutron fluence and neutron flux were obtained and analyzed. In our opinion controversy in the estimation on neutron flux on radiation degradation impact may be explained by presence of the wavelike component in the embrittlement kinetics. Therefore flux effect manifestation depends on fluence level. At low fluencies radiation degradation has to exceed normative value, then approaches to normative meaning and finally became sub normative. As a result of dose rate effect manifestation peripheral RPV’s zones in some range of fluencies have to be damaged to a large extent than situated closely to core. Moreover as a hypothesis we suppose that at some stages of irradiation damaged metal have to be partially restored by irradiation i.e. neutron bombardment. Nascent during irradiation structure undergo occurring once or periodically transformation in a direction both degradation and recovery of the initial properties. According to our hypothesis at some stagem(s) of metal structure degradation neutron bombardment became recovering factor. Therefore self-recovering section of RPV steel radiation embrittlement kinetics is an indication of material smart behavior. As a result oscillation arise that intern lead to enhanced data scatter. Disclosure of the steel degradation oscillating is a sign of the steel structure cyclic self-recovery transformation as it take place in self-organization processes. This assumption has received support through the discovery of the similar “anomalous” data in scientific publications and by means of own additional experiments. Data obtained stimulate looking-for ways to management of the structural steel radiation stability (for example, by means of nano-structure modification for radiation defects annihilation intensification) for creation of the intelligent self-recovering material. Expected results: - radiation degradation theory and mechanisms development, - more adequate models of the radiation embrittlement elaboration, - methods and facility development for usage data of the accelerated materials irradiation for forecasting of their capacity for work in realistic (practical) circumstances of operation, - search of the ways for creating of the stable under neutron irradiation self-recovery smart materials.

The role of chemical composition in WWER RPV irradiation embrittlement

The main chemical elements influencing WWER RPV steel radiation embrittlement are copper, phosphorus, nickel and manganese. Copper and phosphorus are the main influencing elements for WWER-440 RPVs. The influence of copper is implemented in the forming of nanocluster centres rounded by other elements which are the effective barriers to dislocation movements. Phosphorus segregates inside the grains, interacting with matrix defects and attracted to the Cu-type precipitates. Phosphorus also migrates to grain boundaries through diffusion processes. Due to the higher nickel content and very low copper and phosphorus present in WWER-1000 RPV steel, the main influencing element for this RPV type is nickel. The manganese content also has to be taken into account due to synergism of Ni and Mn. The semi-mechanistic model is based on key embrittlement mechanisms of chemical elements influence: matrix damage, irradiation-induced precipitation and element segregation proposed.

Application of Subsize Specimens for Re-Irradiation Embrittlement Monitoring of the First Generation of VVER-440 RPV Steels

Abstract The first generation of VVER-440 Units were not supplied with RPV steel surveillance programs. Therefore, cutting out of samples (templetes) from the inner side of RPV wall is the only possibility for re-irradiation embrittlement monitoring of the first generation of VVER-440 RPV steels. Calculations showed that it is possible to cut out a sample of 7×60×95 mm (with consequent repair) from a RPV core weld. Thus, it is impossible to machine the standard 10×10×55 mm Charpy V-notch specimens to determine RPV steel radiation embrittlement. Therefore it is necessary to use subsize specimens (5×5×27.5 mm and 3×4×27 mm) for ductile-to-brittle transition temperature evaluation of templet material. The small samples were cut out from the inner surface of the all old Russian VVER-440 RPVs to assess the actual condition of the RPV material before and after post-irradiation annealing and after re-irradiation. Special monitoring programs of periodic sampling of RPVs and re-irradiation of the templet material in surveillance channels of VVER-440/213 RPVs were developed. The state of the art of application of subsize specimens for re-irradiation embrittlement monitoring of the first generation of VVER-440 RPV steels is discussed.

Applicability of Sub-Size Charpy Specimens to Transition Behaviour of RPV Materials

The advancement of proper methodology to determination of VVER reactor pressure vessel (RPV) materials transition behaviour has been followed. The project included selection of proper specimen geometries (standard 10x10x55 mm, sub-size 3x4x27 mm and 5x5x27.5 mm Charpy-V samples) and impact as well as dynamic fracture toughness testing. All the fundamental criterions have been applied to consider an applicability of small-sized specimens to Cr-Mo-V and Cr-Ni-Mo-V steel radiation embrittlement studies. Selected specimen geometries have been found to be fully valid for impact and dynamic fracture data and capable for prediction of standard specimen behaviour from the small specimens. Within testing of irradiated materials, it has been found not fully validity of all the compared transition temperatures.

Positron study of radiation embrittlement of steels used in water cooled, water moderated energy reactors

Positron annihilation spectroscopy (PAS) lifetime study was applied in the evaluation of the microstructure parameters and degradation processes of nuclear reactor pressure vessel (RPV) steel surveillance specimens. Study was oriented to the material investigation of Russian WWER-1000 steels (15Kh2MNFAA and 12Kh2N2MAA) with higher Ni content (1.26 wt.% in base metal and 1.7 wt.% in weld). For comparison, the WWER-440 weld metal (Sv10KhMFT) without Ni was measured too. Specimens were studied in as received stage, after irradiation in LVR-15 experimental reactor to the neutron fluence F ( E > 0.5 MeV) = 4 × 10 23 m −2 s −1 and after annealing in vacuum at 475 °C/2 h. Post-irradiation thermal treatment and annealing of defects was well detected by different PAS techniques. It was observed that the sensitivity of PAS parameters to defined irradiation treatment decreases with Ni-content increase. Results confirm the hypothesis that Ni affects size (decrease) and distribution (more homogeneous) of the Cu- and P-rich clusters and M x C x carbides.

Studies of radiation embrittlement of model alloys by thermoelectric measurements and Positron Annihilation Spectroscopy

One of the basic mechanisms of radiation embrittlement of steels is due to matrix damage. Embrittlement results in a raise in the ductile-to-brittle transition temperature, which is used as indicator of the degradation status of the material. Further to mechanical tests, other methods, based on non-destructive measurements, can be also used to follow embrittlement. At JRC-IE measurement of thermoelectric voltage using a specially developed method is regularly performed. The positron annihilation spectroscopy in lifetime set-up is used for study the microstructural changes of matrix due to embrittlement. Both techniques were applied to test several model alloys irradiated at different conditions. Obtained results prove that both techniques are good indicators of the change in material properties.

Use of a Semi-Mechanistic Analytical Model to Analyze Radiation Embrittlement of Model Alloys: Cu and P Effects

We performed analysis of the basic mechanism of radiation embrittlement of steels and weld seams with account of the direct matrix fracture, precipitation and segregation of chemical elements. A model is proposed, which describes matrix fracture due to its neutron bombardment and for 11 model alloys provides an accurate description of the processes of primary embrittlement and re-embrittlement after scheduled heat treatment (annealing). The distinctive features of the proposed model are: a possibility of explaining the embrittlement processes before and after annealing for alloys with low (or zero) Ni content, and its applicability to the analysis of operational behavior of WWER materials.

Cu effects on radiation embrittlement of Ni-containing complex model alloys and the related potentials of the thermoelectric method

One of the major damage contributions to radiation embrittlement of steels is due to copper precipitation. Precipitation mechanism continues until saturation depending on available amount of precipitants, copper concentration in particular. At the JRC-IE measurement of thermoelectric voltage using a specially developed non-destructive method to assess the embrittlement state of materials are regularly performed. The results analysis, from the test of complex sets of nickel containing alloys with different copper content levels, for both fresh and irradiated conditions, are discussed in this paper with the aim of identifying, if copper precipitation to saturation can be followed by non-destructive measurements. The findings are promising showing that the STEAM non-destructive method has very high potential for studying the copper kinetics part in the radiation embrittlement phenomenon.

原子炉圧力容器鋼材(A302B鋼)溶接継手の中性子照射試験

The fast-neutron irradiation embrittlement of structural steels is an important problem in the design of the pressure vessel of a nuclear power reactor. Many studies have been made on the irradiation embrittlement of steels in foreign countries.In Japan, plans are also in progress on the irradiation tests of domestic structural steels, and some have already been carried out by Research Commitee of Irradiation Embrittlement of Steels, The Iron and Steel Institute of Japan, acting as the center's.The present paper shows a part of above studies, the results of irradiation tests with JRR-2 of JAERI and also of the postirradiation tests of domestic steels, namely ASTM A302B steel, aluminium-killed steel and their weld metals.The contents of this paper include (1) irradiation condition of JRR-2, (2) test rig for irradiation, (3) apparatus for the post-irradiation test, and (4) results of the irradiation embrittlement tests and tensile test.The results obtained are as follows:(1) The transition temperature shifts for brittle fracture at 6-9 n/cm2 neutron doses were 67°C on A302B base metal, 107°C on its weld metal and 95°C on its bond metal.(2) The change of tensile property due to neutron irradiation in aluminium-killed steel was larger than in A302B steel.

POSSIBLE FAST-REACTOR CANNING MATERIAL STRENGTHENED AND STABILIZED BY DISPERSION

The high-temperature irradiation embrittlement of steels and nickel alloys is probably due to helium precipitation in grain boundaries. It would be desirable to retain the helium as fine bubbles evenly distributed within the grains. A fine dispersion in the metal matrix may act as nucleation site for helium bubbles and it is proposed to produce by powder-metallurgy techniques afine dispersion of a ceramic oxide in a steel matrix. Ferritic chromium steel was chosen because it is known to be less susceptible to embrittlement than austenitic alloys and in this case the oxide dispersion must also improve the high-temperature mechanical properties.Methods of preparation are discussed and results of preliminary tensile tests performed on iron containing alumina, magnesia, and titania dispersions indicate the feasibility of the proposed solution.