The Relationship of Radiation Embrittlement and Swelling for Austenitic Steels for WWER Internals

Марголин, Б. З.; Kursevich, I. P.; Sorokin, A. A.; Neustroev, V. S.

doi:10.1115/pvp2009-77078

Cited by 7 publications

(4 citation statements)

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“…In the irradiated ASS containing bubbles of Helium and a respectively higher level of void swelling, the chemical changes around the voids led to local transformation into ferrite microstructure. In parallel, the fracture character of the immediate and high T irradiated 18Cr-10Ni-Ti ASSs [34,35] might also indicate the features of DBT behavior. The fractures of the specimens of more than 15-20% void swelling were quasi-cleavage with essentially zero plasticity.…”

Section: Ductile-to-brittle Transitionmentioning

confidence: 98%

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Overview of Intergranular Fracture of Neutron Irradiated Austenitic Stainless Steels

Hojná

2017

Metals

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Austenitic stainless steels are normally ductile and exhibit deep dimples on fracture surfaces. These steels can, however, exhibit brittle intergranular fracture under some circumstances. The occurrence of intergranular fracture in the irradiated steels is briefly reviewed based on limited literature data. The data are sorted according to the irradiation temperature. Intergranular fracture may occur in association with a high irradiation temperature and void swelling. At low irradiation temperature, the steels can exhibit intergranular fracture at low or even at room temperatures during loading in air and in high temperature water (~300 • C). This paper deals with the similarities and differences for IG fractures and discusses the mechanisms involved. The intergranular fracture occurrence at low temperatures might be correlated with decohesion or twinning and strain martensite transformation in local narrow areas around grain boundaries. The possibility of a ductile-to-brittle transition is also discussed. In case of void swelling higher than 3%, quasi-cleavage at low temperature might be expected as a consequence of ductile-to-brittle fracture changes with temperature. Any existence of the change in fracture behavior in the steels of present thermal reactor internals with increasing irradiation dose should be clearly proven or disproven. Further studies to clarify the mechanism are recommended.

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Section: Ductile-to-brittle Transitionmentioning

confidence: 98%

“…The tensile test The other example shows the behavior of Russian 18Cr-10Ni-Ti steel [34] (Figure 12). The fracture appeared mostly IG after the irradiation in BOR-60 at 400-450 • C and ruptured in tensile tests at RT and 290 • C. The steel contained 3-13% void swelling.…”

Section: High Irradiation Temperaturementioning

confidence: 99%

Overview of Intergranular Fracture of Neutron Irradiated Austenitic Stainless Steels

Hojná

2017

Metals

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“…These phases are generally Ni-Si rich clusters and are believed to form because of radiation induced segregation of Ni and Si to sinks (dislocations, grain boundaries, voids surfaces etc.). In addition to these common phases some studies showed that ferrite phase also may form under irradiation [49,50] but due to scarcity of experimental data we exclude them from current work. In the sections below we first discuss the precipitation of carbides and then the γꞌ and G-phase.…”

Section: Kineticsmentioning

confidence: 99%

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

et al. 2017

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The current work combines the Cluster Dynamics (CD) technique and CALPHAD-based precipitation modeling to address the second phase precipitation in cold-worked (CW) 316 stainless steels (SS) under irradiation at 300-400 °C. CD provides the radiation enhanced diffusion and dislocation evolution as inputs for the precipitation model. The CALPHAD-based precipitation model treats the nucleation, growth and coarsening of precipitation processes based on classical nucleation theory and evolution equations, and simulates the composition, size and size distribution of precipitate phases. We benchmark the model against available experimental data at fast reactor conditions (9.4 × 10 -7 dpa/s and 390 °C) and then use the model to predict the phase instability of CW 316 SS under light water reactor (LWR) extended life conditions (7 × 10 -8 dpa/s and 275 °C). The model accurately predicts the γꞌ (Ni 3 Si) precipitation evolution under fast reactor conditions and that the formation of this phase is dominated by radiation enhanced segregation. The model also predicts a carbide volume fraction that agrees well with available experimental data from a PWR reactor but is much higher than the volume fraction observed in fast reactors. We propose that radiation enhanced dissolution and/or carbon depletion at sinks that occurs at high flux could be the main sources of this inconsistency. The integrated model predicts ~1.2% volume fraction for carbide and ~3.0% volume fraction for γꞌ for typical CW 316 SS (with 0.054 wt.% carbon) under LWR extended life conditions. This work provides valuable insights into the magnitudes and mechanisms of precipitation in irradiated CW 316 SS for nuclear applications.

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“…Reliable data of voids swelling have been obtained in conditions of different neutron spectrum -in fast reactors or accelerators [1], [2]. Data shows 20 % and higher void swelling presence in the steels results in a drastic and total loss of strength and plasticity, but the findings were obtained for the steels irradiated up to 72 dpa at fast neutron spectrum (BOR-60 reactor) at about 400-550˚C.…”

Section: Introductionmentioning

confidence: 99%

Estimation of material degradation of VVER-1000 baffle

Harutyunyan¹,

Košál²,

Vandlik³

et al. 2017

EPJ Web Conf.

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Abstract. The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor.

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The Relationship of Radiation Embrittlement and Swelling for Austenitic Steels for WWER Internals

Cited by 7 publications

References 0 publications

Overview of Intergranular Fracture of Neutron Irradiated Austenitic Stainless Steels

Overview of Intergranular Fracture of Neutron Irradiated Austenitic Stainless Steels

Integrated modeling of second phase precipitation in cold-worked 316 stainless steels under irradiation

Estimation of material degradation of VVER-1000 baffle

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