PSI - Issue 17

Salatiel Pérez Montejo et al. / Procedia Structural Integrity 17 (2019) 123–130 Salatiel Pérez Montejo et al/ Structural Integrity Procedia 00 (2019) 000 – 000

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1. Introduction

The beltline zone of a BWR vessel (Fig. 1) is adjacent to the nuclear core. As a result, the material is under radiation and it is embrittled, when operation hours are accumulated. Besides, different loading conditions take place during the operation life (USNRC, 1988). Surveillance programs have been developed in order to determine the fracture toughness of the material at any time. The data obtained is useful in the evaluation of the structural integrity of the vessel and its internals. In this way, the probability of a failure is reduced during the normal and emergency conditions of operation, such as a Small Loss of Coolant Accident (SLOCA) or a Medium Loss of Coolant Accident (MLOCA). (Qian, 2013. Qian, 2014)

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Fig. 1. (a) A BWR-5 vessel; (b) Beltline zone.

The Nuclear Regulatory Commission (NRC) has presented the following criteria, mainly for the evaluation of the effects occasioned by an Extended Power Uprate (USNRC, 2003). (1) The 10 CFR 50 Appendix H establishes the surveillance program of the materials of BWR and PWR. In this way, the degradation of the fracture toughness of a ferritic steel at the beltline zone is monitored. Test coupons are irradiated in reactor surveillance capsules to facilitate the evaluation of vessel fracture toughness. They are periodically extracted from the reactor. In this way, the decay of this fracture parameter can be evaluated (USNRC, Appendix A, 2017); (2) The 10 CFR 50 Appendix G specifies the fracture toughness requirements of ferritic materials of PWR and BWR. In this way, adequate safety margins can be established in any normal operating condition, as well as anticipated operational incidents . A pressure-temperature limit can be established during the useful life of the reactor. The ASME code is the basis of the requirements of this appendix (ASME, 2007); and (3) The 10 CFR 50.60 establishes the requirements for the compliance with the requirements of the 10 CFR 50 appendix H. The USNRC governs the process of the renewal of the operating license of a nuclear power plant in the United States. Once its Commercial Operation License has expired, its license must follow the process called License Renewal that is regulated by the 10 CFR 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants (Ruiz López. P, 2017). In the evaluation of the renewal of the operating license, it has to be demonstrated that the passive components and those related with security must maintain their structural integrity during the extended period of operation. Under these conditions, a reactor vessel has to operate during sixty years.

Nomenclature ASME

American Society of Mechanical Engineers

CFR EPU FEM

Code of Federal Regulations Extended Power Uprate Finite Element Method