On extending the life of nuclear reactors

To extend the service life of nuclear reactors, witness samples from the shells of the core of the reactor vessel are placed in their core. According to the requirements in force in the industry, the reference samples are loaded into the reactor plant unloaded up to the design stresses. This can lead to a biased assessment of the possible extension of the reactor’s life. In connection with the above, in order to assess the mutual influence of operating factors and the stress-strain state of the base metal and welded joints on embrittlement, the reference specimens must be loaded with a force that causes the maximum possible stresses in the specimens during the operation of the reactor. On the basis of domestic and international experience, a test procedure, design and loading scheme for compact witness samples are proposed for modeling and assessing the mutual influence of operating factors and stress-strain state on the object under study (VVER power reactor vessel). For VVER RPVs, the duration of the additional service life should be confirmed by the justification that by the end of the additional service life, the fracture toughness values of the base metal and metal of the welded seams located in the irradiation zone will allow without destruction to withstand all operational and emergency loads, as well as loads at hydraulic tests.

Evaluation of a sensitive, reasonable, and fast detection method for 55Fe in steel

A pilot study to quantify 55Fe in steel from a reactor vessel of a nuclear power plant by accelerator mass spectrometry (AMS) without any chemical sample preparation was validated by liquid scintillation counting (LSC) and AMS after radiochemical separation. AMS reaches an uncertainty < 10% at the 1 kBq gFe−1 level within less than 10 min measuring time. The background was < 3 Bq gFe−1, presently limited by the short measurement time. The new instrumental AMS method for analysing 55Fe from neutron capture production is reasonable and fast compared to other analytical methods.

Finite Element Welding Residual Stress Analysis of CRDM Penetration Nozzles

This article investigates numerically welding residual stress distributions of a tube with J-groove weld in control rod drive mechanisms of a pressurized nuclear reactor vessel. Parametric study is performed for the effect of the tube location, tube dimensions, and material's yield strength. It is found that residual stresses increase with increasing the inclination angle of the tube, and the up-hill side is the most critical. For thicker tube, residual stresses decrease. For material's yield strength, both axial and hoop residual stresses tend to increase with increasing the yield strength of Alloy 600. Furthermore, axial stresses tend to increase with increasing yield strength of Alloys 82/182.

Analysis of in-vessel accident progression in VVER1000 NPP during SBO accident with external reactor vessel cooling method

In this study, the MELCOR v1.8.6 code was utilized to perform an analysis of the in-vessel accident progression in VVER1000 reactor during the Station Black-Out (SBO) accident with and without external reactor vessel cooling (ERVC) strategy. The analysis presented the predictions of the main phenomena during the accident such as failure of fuel cladding, collapse of lower core support plate, relocation of core debris to lower plenum and mass of debris components in lower plenum, and provided comparisons between two cases in term of main parameters such as integrity time of reactor and structure components of molten pool. These parameters are very important inputs for further research on the application of external vessel cooling strategy for VVER1000 reactor.

Study on Buckling Strength and Post Buckling Behaviors of Reactor Vessel Lower Heads

As the lessons learned from the Fukushima Daiichi Nuclear Power Plant accident, the importance of controlling the behavior after a failure and mitigating consequences of a failure was recognized. Conventional reactor structural design has been aimed at preventing the occurrence of failure due to Design Basis Events (DBE). This study aims to improve the resilience of the reactor structure under Beyond Design Basis Events (BDBE), such as very high temperatures and excessive earthquakes during severe accidents, by mitigating the consequences after failure.