4.05.1 Introduction

The ferritic steel reactor pressure vessel (RPV) of a light water reactor (LWR) is unique in terms of nuclear plant safety. This is because the RPV is a pressure boundary component whose catastrophic failure by brittle fracture could lead to severe core damage and, potentially, to the widespread release of radioactivity. In addition to the thermal, mechanical, and chemical degradation processes common to all primary circuit components, the ferritic steel RPV, because of its close proximity to the reactor core, also undergoes changes in mechanical properties because of radiation damage from the flux of fast neutrons arising from nuclear reactions in the fuel. The integ­rity of the RPV has to be assured throughout the life of the plant. Irradiation results in hardening and embrittlement processes, the most important effect of which is the rise in the ductile-brittle transition temperature (DBTT) and the decrease in the frac­ture toughness of the RPV. This can be an important issue in the region of the beltline which experiences the highest neutron fluence. These changes in mechanical properties make the RPV more suscepti­ble to brittle fracture as the plant ages. The primary purpose of this chapter is to demonstrate our current understanding of such radiation damage effects in ferritic pressure vessel steels.

Low alloy ferritic steel pressure vessels are employed in all western LWRs (both boiling water reactors (BWRs) and pressurized water reactors (PWRs)), and also in VVER (LWR) reactors in Russia and a number of European countries (see, e. g., Steele and Sterne1). In the past, ferritic pressure vessel steels were also employed in gas-cooled Magnox reactors in the United Kingdom2 (Magnox reactors with steel vessels are now undergoing decommissioning ).

Demonstrating the safe operation of such a plant has led to extensive international research over the last 40-50 years on the aging effects in ferritic steels. The need for such scientific understanding has been raised at the highest level. For example, Sir Alan Cottrell, then UK Government Chief Scientist, in a memorandum concerning the integrity of LWR pressure vessels, dated 22 January 1974, to the UK Parliament Select Committee on Science and Indus­try stated ‘‘The possible gradual growth of small cracks in highly stressed regions, by ageing and corro­sion effects during service needs further scientific

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investigation.

The discussion here establishes that in the field of radiation damage, it is the direct application of fun­damental research to operating reactors that is signif­icant. This chapter demonstrates that developments in the understanding ofthe damage mechanisms have enabled an improved description of the in-service properties of RPVs of operating reactors. Indepen­dent peer review has been central to the process and particularly important from a regulatory perspective. Frequently, it is not simply the research community directly involved that has to assess any improved description ofthe degradation processes; for example, the safety authorities within the utility, operating the reactor (or fleet), or the national regulator will become involved in the process.

Most importantly it has not always been possible to predict the end of life (EOL) vessel properties from the data obtained from materials irradiated as part of vessel surveillance programs. The vessel sur­veillance programs for commercial nuclear reactors are intended to monitor the irradiation-induced changes in mechanical properties of life-limiting structural materials subjected to significant neutron fluence. Thus, they are designed to provide advance information concerning the state of degradation in the mechanical properties of key structural compo­nents. However, because of the inevitable differences in neutron dose rate between the vessel wall and such surveillance samples (typically a factor of5 or more), the scarcity of such data at the start of plant opera­tion, and complex and unexpected embrittlement dependencies on steel composition, it has become necessary to develop dose-damage relationships (DDRs) on the basis of mechanistic understanding that predict the embrittlement dependence on mate­rial and irradiation variables.

The vast majority of investigations on the aging effects in ferritic steels over the last 40-50 years have been concerned with the effects of neutron irradiation over quite a narrow set of irradiation conditions, for example, irradiation temperatures of ^250-300 °C (although there was interest in irradiation tempera­tures as low as 160 °C) and irradiation doses typical of in-service exposures (<0.1 dpa). Early studies focused, primarily, on the effects of irradiation on mechanical properties, while in the last 20 years such studies have been combined with rigorous investigations of the effects ofmaterials and irradiation variables on the microstructure developed under irradiation. It is this combination that has allowed significant insight into the mechanisms that determine the effect of radiation damage on the bulk properties of ferritic RPV steels. As will be shown, such studies have encompassed simple model alloys, steels with a controlled composi­tion, and commercial steels employed in real vessels. A major outcome in recent years is the development of mechanistically based or guided DDRs that are employed, frequently in a regulatory framework, to predict the behavior of reactor vessels. (Some work­ers refer to DDRs as ‘embrittlement correlations’.) In formulating such DDRs, it has been necessary to encompass not only the irradiation variables such as flux, fluence, and irradiation temperature but also material factors, such as composition, heat treatment, and product form.

In reviewing such a long-standing field, it is neces­sary to provide a focus to enable presentation of a manageable set of data. As stated earlier, the chapter focuses on the mechanisms that control RPV em­brittlement, how such understanding has been incorporated into mechanistically based DDRs, and the limitations or current research issues associated with their development. These mechanistically based DDRs have been developed primarily for LWRs and Magnox reactors located in the West and this chapter gives prominence to such studies; that is, the chapter does not provide an extensive review of studies on the embrittlement of the RPVs of VVER reactors (see Nikolaev et al.,5 Kryukov et al,6 Shtrombakh and Nikolaev,7 and Brumovsky8 for such a review). Its organization reflects the overall focus; first, the chapter provides a description of RPV steels (Section 4.05.2) and then provides an overview of the effects of radiation damage on their mechan­ical properties (Section 4.05.3). Section 4.05.4 describes the status of our mechanistic understanding of embrittlement caused by radiation damage. This understanding has proved pivotal to the development of the DDRs described in Section 4.05.5. Outstand­ing issues, particularly those arising through plant life extension, are outlined in Section 4.05.6.