The Studies of Irradiation Hardening and RIS on IASCC of Reactor Internals Bolts

IASCC of stainless steel has been the most important issue for internals BFBs. The inspection data analysis indicates that there is a closed relation between irradiation fluence and cracked BFBs distribution. Then the nanoindentation and 3DAP tests were carried out to study the hardening and radiation induced segregation (RIS) behaviors of the reactor internals stainless steel specimens irradiated with 6 MeV Xe ions at room temperature. It is indicated that higher irradiation damage will cause more significant hardening and RIS and consequently increase the IASCC susceptibility.

Investigation of irradiated metal of WWER-type reactor internals after 45 years of operation. Part 1. Research program and cutting out of samples from pressure vessel internals

The program is presented for investigations of the metal of the most irradiated elements of the WWER-440 reactor of the Novovoronezh NPP Unit 3 decommissioned after 45 years of operation. The fragments (cylindrical samples) were cut out from various zones of the core baffle and segment of forming ring of core barrel.

Investigation of irradiated metal of WWER-type reactor internals after 45 years of operation. Part 3. Microstructure and phase composition

TEM, SEM, and APT techniques have been used to analyze radiation-induced components of metal structure of fragments cut from the pressure vessel internals of Novovoronezh NPP Unit No 3 after 45 years of operation. The fragments differed in the neutron damaging doses (from 14 to 43 dpa) and the irradiation temperature (from 285 to 315°C). The density and dimensions of titanium carbides and carbonitrides, dislocation loops, radiation-induced voids, segregations, and nanoscale precipitates were determined. The contributions of structural components to the radiation hardening of the investigated fragments of 18Cr-10Ni-Ti stainless steel were estimated.

Improved Computational Fluid Dynamics Framework for Reactor Core Baffle Swelling Assessment

This paper presents an improved estimation of reactor core baffle temperature distribution, during operation, at the nominal power level to address swelling problems of the reactor internals. Swelling is the main limiting factor in the reactor core internals long term operation of VVER-1000 nuclear units. The material irradiation-induced swelling and creep models are very sensitive to temperature distribution in metal; thus, a more detailed analysis of the core baffle metal thermohydraulic cooling characteristics is required. A framework for the computational fluid dynamics (CFD) analysis of VVER-1000 reactor baffle cooling is presented. First, an analytical model was developed to obtain boundary conditions (BCs) and simplify CFD analysis. Second, the CFD analysis was performed using 60 deg symmetry, which included core, baffle, and core barrel, and it is limited by the height of the baffle. Core is simplified as an equivalent coolant domain with considering of spatial volumetric energy release. Core baffle is presented as monolithic body with considering of gamma-ray heat generation. Model includes cooling ribs and simplified geometry of connecting studs, with cooling flow of the coolant through the nuts grooves. Calculated convection coefficient and temperature are in good agreement with analytical model and give a more accurate result comparing to RELAP5/mod3.2. Obtained temperature field was used to estimate baffle swelling process and justify safe long term operation of the reactor internals.