Radiation-induced degradation of stainless steel light water reactor internals

Kenik, E.A.; Busby, Jeremy T.

doi:10.1016/j.mser.2012.05.002

Cited by 102 publications

(36 citation statements)

References 65 publications

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“…All immersion density measurements and TEM observations have shown that swelling depends on the metallurgical state (cold-worked austenitic stainless steel (CW) versus solutionannealed (SA) one) [1][2][3]. Using the conventional bright fieldand dark field-TEM imaging technique in combination with the energy filtered-and high resolution-TEM imaging technique (CBF-, CDF-, EF-and HR-TEM, respectively), it has become possible to carry out fuller investigations on precipitation behavior under neutron irradiation [4,5].…”

Section: Introductionmentioning

confidence: 99%

Correlation of radiation-induced changes in microstructure/microchemistry, density and thermo-electric power of type 304L and 316 stainless steels irradiated in the Phénix reactor

Laborne

Gavoille

Malaplate

et al. 2015

Journal of Nuclear Materials

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Section: Introductionmentioning

confidence: 99%

Correlation of radiation-induced changes in microstructure/microchemistry, density and thermo-electric power of type 304L and 316 stainless steels irradiated in the Phénix reactor

Laborne

Gavoille

Malaplate

et al. 2015

Journal of Nuclear Materials

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confidence: 99%

Microstructural Evolution of High-Strain-Rate Severe Plastic Deformation Processed 316L during Kr Ion Irradiation and elevated Temperature Exposures

et al. 2016

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Austenitic stainless steels (SS) are used as structural materials in reactor internal components of nuclear power plants and also in fast reactors [1]. They degrade by prolonged elevated temperature exposure to neutron irradiation, leading to radiation enhanced kinetics for point defect migrations, manifest in hardening, embrittlement and void swelling, irradiation creep, radiation-induced segregation (RIS) and precipitation [1,2]. RIS accelerated thermally-driven phase transformations are expected to become increasingly relevant for the degradation of austenitic SS under extended life conditions and development of materials with improved irradiation-damage tolerance is desirable [2]. Enhanced irradiation damage tolerance has been reported for materials with microstructures comprising high densities of sink sites for annihilation of excess point defects, e.g. intra-granularly provided by thermodynamically stable precipitates and inter-granularly as grain boundaries [3][4][5][6][7].Here we report a transmission electron microscopy (TEM) based investigation of the microstructural development of 316L SS alloys modified by novel high-strain rate severe plastic deformation (SPD) methods based on linear plain strain processing under exposures to Kr ion-irradiation and elevated temperatures [8][9][10][11]. The SPD-modified SS offer significantly improved mechanical strength, fatigue performance, resist thermal coarsening up to ~650˚C, and exhibit highly deformed, ultra-fine-grain to nano-scale refined grains of austenitic-Fe with lath-like morphology and trace amounts of martensite [8][9][10][11] (Fig. 1). In-situ irradiation TEM experiments were performed with 1MeV Kr ions at 298K, 573K, 673K and 773K for fluence up to ~1x10 17 ions/cm 2 , equivalent to about 20dpa, using the IVEM-Tandem Facility at Argonne National Laboratory. In-situ sequences during heating and ion-irradiation have been recorded typically with a frame rate of 10 frames per second with 300keV electrons using a double-tilt heating holder. TEM analyses prior to and after the ion-irradiation exposures have been performed using a Jeol JEM2100F and a FEI Tecnai G2 F20ST operated at 200kV. Experimental details for automated crystal orientation mapping (ACOM), which has been performed with NanoMegas ASTAR and TOPSPIN systems, and for TEM specimen preparation have been reported previously [8][9][10][11]. The 316L SS microstructure remained austenitic, resisted significant grain coarsening, retained its morphology and did not exhibit any precipitation of intermetallic or carbide phases after up to ~6.5h of Kr ion irradiation (~ 20dpa) at up to 773K (e.g. Fig. 2). In-situ TEM revealed the formation of interstitial dislocation loops and interactions of the irradiation induced defects with each other and the pre-existing intra-granular dislocation structures and the grain boundaries, leading to reduction in dislocation density and local grain disorientations. The dynamic in-situ TEM observations enabled identification of different fluence dependent regimes duri...

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“…Consequences of these microstructural and microchemical evolutions are increased hardening, embrittlement, enhanced rates of subcritical cracking, low-temperature irradiation creep, void swelling, loss of ductility and creep resistance [1][2][3][4].…”

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An intermetallic forming steel under radiation for nuclear applications

Hofer

Stergar

Maloy

et al. 2015

Journal of Nuclear Materials

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In this work we investigate the formation and stability of intermetallics formed in maraging steels under ion beam radiation utilizing nanoindentation, microcompression testing and atom probe tomography. A comprehensive discussion analyzing the findings utilizing rate theory is introduced, comparing the aging process to radiation induced diffusion. New findings of radiation induced segregation of undersize solute atoms (Si) towards the precipitates are considered.Structural materials in nuclear applications suffer a wide range of microstructural changes under radiation. Consequences of these microstructural and microchemical evolutions are increased hardening, embrittlement, enhanced rates of subcritical cracking, low-temperature irradiation creep, void swelling, loss of ductility and creep resistance [1][2][3][4].The formation of cascades and the formation of a large number of point defects due to particle interaction with materials cannot be avoided initially. However, recent studies have shown that the collection of point defects into voids or clusters can be delayed or avoided by trapping the initial defects at a large number of defect sinks where the defects then annihilate, preventing their accumulation within the lattice [5][6][7][8]. The most effective defect sinks and traps have been proven to be interfaces such as grain boundaries or phase boundaries [5][6][7][8]. This information has led to the creation of materials with extremely high interface densities which allow for a more radiation tolerant system by utilizing a high density of nm-sized precipitates or multilayer structures [9][10][11]. The influence of these precipitates or multilayers on the mechanical properties and radiation resistance of the material is controlled by parameters like size, shape, number density or layer thickness and volume fraction. One well known example of this approach are Nanostructured Ferritic Alloys (NFAs), where extremely large numbers of nanoscale oxide particles (in the range of 10 23 -10 24 m -3 in number density) are present. These materials are also known as mechanically alloyed oxide dispersion strengthened (ODS) steels [12]. Another example is the widely studied immiscible fcc/bcc metallic 3 multilayers such as Cu/Nb multilayers realized by physical vapor deposition (PVD) coating techniques or roll bounding [13,14]. This microstructure allows the radiation induced point defects or nuclear reaction products to "heal" or become trapped at interfaces. These materials are usually considered top down structures where the manufacturer creates these materials from powders or sputter targets and consolidates those during processing. This leads to the fact that the resulting microstructure can rarely be changed after materials synthesis. NFAs after processing into tubes or thin plates have been shown to be strongly textured with anisotropic properties formed during processing while the Cu/Nb layers are by nature anisotropic. Post processing heat treatment usually does not change the structure any more due to the fact t...

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Radiation-induced degradation of stainless steel light water reactor internals

Cited by 102 publications

References 65 publications

Correlation of radiation-induced changes in microstructure/microchemistry, density and thermo-electric power of type 304L and 316 stainless steels irradiated in the Phénix reactor

Correlation of radiation-induced changes in microstructure/microchemistry, density and thermo-electric power of type 304L and 316 stainless steels irradiated in the Phénix reactor

Microstructural Evolution of High-Strain-Rate Severe Plastic Deformation Processed 316L during Kr Ion Irradiation and elevated Temperature Exposures

An intermetallic forming steel under radiation for nuclear applications

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