Development and application of safety assessment system in PFM analysis for RPV to study the influence of the neutron fluence estimate. (November 2018)

• Record Type: Journal Article
• Authors: Chang, Ching-Hsin; Chou, Hsoung-Wei; Huang, Chin-Cheng
• Is Part Of: International journal of pressure vessels and piping. Volume 167(2018)
• Page Start: 43
• Page End: 51
• Publication Date: 2018-11
• Subjects: Probabilistic fracture mechanics -- Reactor pressure vessel -- Radiation embrittlement -- Neutron fluence -- Programming; Pressure vessels -- Periodicals; Pipe -- Periodicals; Récipients sous pression -- Périodiques; Tuyaux -- Périodiques; Pipe; Pressure vessels; Periodicals; 681.76041
• DOI: 10.1016/j.ijpvp.2018.10.007
• Languages: English
• ISSNs: 0308-0161

Abstract

Abstract: The reactor pressure vessel (RPV) is the most important core container for the power generation system of a nuclear power plant. However, the vessel shell is prone to radiation embrittlement because of the operational environment of the power generation process which may lead to crack growth in the shell wall. For the purpose of estimating and ensuring the integrity of the RPV structure, the risk to and reliability of the structure after a period of operation needs to be analyzed under a variety of transient loading conditions. Probabilistic fracture mechanics (PFM) is a new method of analysis applied in a number of studies for estimating the safety of component structures. However, it is difficult for power plant technicians and engineers to evaluate the influence of circumstances arising under various accidental situations on the RPV by PFM analysis because of the lack of relevant experience. Therefore, this study first develops an analysis system which integrates the Oak Ridge Nuclear Laboratory's (ORNL's) PFM FAVOR computer code and analysis models of Taiwan's domestic RPVs. This system may offer engineers and researchers the ability to estimate and analyze the risk and reliability of both boiling water reactors and pressurized water reactor vessels quickly and correctly. Secondly, the system is applied to study the influence of the neutron fluence and sampling conditions on the rupture probability of the RPV at the end of license including 36 and 54 effective full power years (EFPY). The four types of methods used for carrying out the PFM analysis are the Extrapolation, Average, NLR2 and NLR1 methods, and four representative pressurized thermal shock (PTS) transients were considered as loading conditions. The results show that a lower rupture probability is obtained when using the non-linear regression methods to predict the neutron fluence and assess the RPV integrity. Furthermore, the difference in the resulting values will rise with increasing the operation period. Under the same conditions, the rupture probability sequence by size is Average > Extrapolation > NLR2 > NLR1, espacially at 54 EFPY.