Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

doi:10.1016/j.jnucmat.2003.11.007

Journal of Nuclear Materials

Volume 326, Issue 1, 1 March 2004, Pages 19-29

https://doi.org/10.1016/j.jnucmat.2003.11.007 Get rights and content

Abstract

Irradiation damage in three austenitic stainless steels, SA 304L, CW 316 and CW Ti-modified 316, is investigated both experimentally and theoretically. The density and size of Frank loops after irradiation at 320 and 375 °C in experimental EBR II, BOR-60 and OSIRIS reactors for doses up to 40 dpa are characterized by TEM. The evolution of the initial dislocation network under irradiation is evaluated. A cluster dynamics model is proposed to account quantitatively for the experimental findings.

Introduction

The internal structures of pressurized water reactors (PWR) located close to the reactor core are used to support the fuel assemblies, to maintain the alignment between assemblies and the control bars and to canalize the primary water. In general these internal structures consist of baffle plates in Solution Annealed 304 stainless steel and baffle bolts in Cold Worked 316 stainless steel. These components undergo a large neutron flux at temperatures between 280 and 380 °C. As a result, the material exhibits a substantial increase in yield stress and reduction in ductility which may be damaging to the proper operation of the reactor. For instance the observed cracks in bolts, usually attributed to irradiation assisted stress corrosion cracking (IASCC), can be seen as a consequence of the evolution of plasticity in these materials loaded in a corrosive medium, together with possible evolutions of grain boundary chemistry (Radiation Induced Segregation) which is well documented [1]. This irradiation induced embrittlement becomes more pronounced with increasing irradiation dose, thus being an important and limiting factor in ageing reactors. Of special interest is the question of the possible saturation of microstructural evolution and related hardening for long term irradiation. In the present contribution, we will focus our attention on the evolution of the structural defects induced by irradiation and on their consequences on the yield stress, which we acknowledge as being only one of the contribution to embrittlement phenomenon. This evolution in mechanical properties depends both on the initial metallurgical state of the alloy and on the irradiation conditions (temperature, flux, dose, and neutron energy spectrum). It is associated with a microstructural evolution resulting from the production and the collective dynamics of irradiation point defects, leading to the formation of dislocation loops and cavities, and to the evolution of the initial dislocation network. In order to rationalize the effects of irradiation conditions, and to predict the long term behavior from the observations for smaller doses, it is necessary to develop models able to describe, on one side the microstructural evolutions, and on the other side the resulting hardening effects. The parameters entering these models have to be identified from a quantitative description of the microstructure and yield stresses, and the prediction of the models have to be validated by comparison with experiments for larger dose. The aim of the two companion papers is to provide with quantitative characterization of irradiation defects by TEM (paper I) and of mechanical properties (yield stress) after irradiation by tensile test at constant strain rates (paper II). The modeling of microstructure evolution will be done using a cluster dynamics approach (paper I) whereas the evolution of yield stress for a given microstructure will be modeled using classical dislocation theory (paper II).

The two papers are structured as follows: in paper I, we present the materials investigated, the irradiation conditions and the characterization methods (Section 2). The experimental results concerning the microstructures for the different alloys and irradiation conditions will be presented in Section 3. The principles of the cluster dynamic model, the identification of the parameters and the predictions are developed and compared with experimental results in Section 4. In paper II, Section 1 recalls the proposed mechanism for the evolution of yield stress, and summarizes briefly and qualitatively the findings of paper I, whereas Section 2 presents the experimental conditions for tensile testing and the measured yield stresses for the different alloys and irradiation conditions. In Section 3 we propose a model for the yield stress evolution coupling the hardening by dislocation loops and the question of their stability, and we compare the predictions of this model with the experimental results.

Section snippets

Materials

The alloys for the microstructural characterization after neutron irradiation are 300 series stainless steels commonly used in PWR internals: Solution Annealed 304L (used for baffle plates) and Cold Worked 316 (used for baffle bolts). In addition, an exploratory material such as modified 316 containing a small amount of Ti, has been investigated. The chemical compositions of the three alloys are provided in Table 1. The choice of the alloys was motivated by their relevance in actual reactors.

Experimental results

The experimental results for the three alloys are presented for each irradiation temperature. We present first the microstructure after irradiation at the intermediate temperature (330 °C) obtained either in the mixed flux reactor OSIRIS at low irradiation doses or in the fast breeder BOR-60 reactor for the higher irradiation doses. We then give the results concerning the irradiation at the highest temperature (375 °C) in the fast breeder reactor EBR-II.

Modeling

The physical reasons for irradiation damage stems from the collective behavior of the point defects created during irradiation. For a given neutron flux, the number of vacancies and interstitials, and more generally, the number of groups of point defects effectively created after a collision depends on the energy of the incoming neutron, and on the events taking place at the atomistic level in the irradiation cascade. There exists a considerable amount of simulation work, mainly using molecular

Conclusions

A systematic experimental quantitative characterization of irradiation induced microstructures in austenitic stainless steels has been performed for different irradiation conditions in terms of temperature, fluxes, doses and energy spectra. These results have been analyzed using a cluster dynamics model specially adapted to account for the evolution of dislocation substructure, and to deal with large doses.

All the parameters in the model have a transparent physical interpretation. The values of

Acknowledgements

Enlightening discussions with Dr G. Martin, Dr P. Bellon, Dr R. Cauvin and Dr Ch. Domain are gratefully acknowledged. Authors are also grateful to Dr G.R. Imel at ANL, US, for performing neutron irradiation in EBR-II reactor and to Dr V. Golovanov and Pr V. Shamardin at RIAR, Russia, for performing neutron irradiations in BOR-60 reactor. This work was performed in the frame of the French R&D Project ‘PWR Internals’ sponsored by Electric Power Research Institute (MRP/Joint Owner Baffle Bolts

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Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

Abstract

Introduction

Section snippets

Materials

Experimental results

Modeling

Conclusions

Acknowledgements

J. Nucl. Mater.

J. Nucl. Mater.

J. Nucl. Mater.

J. Nucl. Mater.

J. Nucl. Mater.

J. Nucl. Mater.

J. Nucl. Mater.

Mater. Res. Soc. Symp. Proc.