The primary heat transport system consists mainly of the in-core fuel channels connected to the steam generators by a system of feeder pipes and headers. The feeders and headers are made of carbon steel. Feeders run vertically upwards from the fuel channels across the face of the reactor and horizontally over the refueling machine to the headers. Structural materials of the primary systems of nuclear power plants (NPPs) are exposed to high temperature and pressure conditions, so that the materials employed in these plants have to take into accounts a useful design life of at least 30 years. The corrosion products, mainly iron oxides, are generated from the carbon steel corrosion which is the main constituent of the feeder pipes and headers of this circuit. Typical film thickness on CANDU-PHWR surface is 75μm or 30mg/cm2. Deposits on PHWR tends to be much thicker than PWR due to use of carbon steel and also for the source of corrosion products available on the carbon steel surface. Degradation of carbon steel for the feeder pipes transferring the primary system coolant by flow-assisted corrosion in high temperature has been reported in CANDU reactors including Point Lapreau, Gentully-2, Darlington and Bruce NPPs. The formation of Fe3O4 film on a carbon steel surface reduces the dissolution rate of steel substantially. The protectiveness of the Fe3O4 film over the carbon steel is affected by the environmental factors and the operational parameters of the feeder pipes, including the velocity, wall shear stress, solution pH, temperature, concentration of dissolved iron, quality of solution, etc. For effective chemical decontamination of these thick oxides containing radionuclides such as Co-60, it is necessary to understand the corrosion behaviors of feeder pipes and the characteristics of oxide formed on it. In this work, we investigated the growth of oxide films that develop on type SA-107 Gr. B carbon steel in high temperature water and steam environment by scanning electron microscopy (SEM) and glow discharge optical emission spectrometry (GD-OES) for the quantification and the solidstate speciation of metal oxide films. This study was especially focused to set the experimental tests conditions how to increase the oxide thickness up to 50 m by changing the oxidation conditions, such as solution chemistry and thermo-hydraulic conditions both temperature and pressure and so on.

Canada’s Pickering Unit 3 was performed a three-stage decontamination from June to August 1989 in preparation for pressure tube replacement. The first step was a reducing CAN-DECON treatment to dissolve the magnetic film inside the reactor, which was applied following partial defueling of the reactor core. The second step was an oxidative dilute alkaline permanganate treatment to remove the chromium-rich oxides of the stainless steel parts. And the final CAN-DECON step was applied continuously after completely removing fuel from the reactor core. In situ pipe gamma-ray spectroscopy techniques were applied to measure radioactivity within feeder piping during various stages of Pickering Unit 3 decontamination. Measurements were performed at a maximum dose rate of 5 mSv/h, and both the detector and the scanned feeder pipe were properly shielded from other neighboring pipes. 60Co was the dominant radionuclide in feeder piping prior to decontamination. And radionuclides 103Ru, 95Zr, 95Nb, 59Fe, 140La and 124Sb were detected. The Co-60 radioactivity was 2.09×105 Bq/cm2 before decontamination and 3.11×103 Bq/cm2 after decontamination in the inlet feeder pipe T18. And in the outlet feeder pipe P21, it is 2.56×104 Bq/cm2 before decontamination and 2.04×103 Bq/cm2 after decontamination.

The Heavy Water Reactor(HWR) Heat Transport(HT) system transient analysis for the design of major nuclear equipment during normal and abnormal operating conditions was performed. The compliance with requirements of AECB Regulatory Document R-77 for CANDU reactor was estimated in CANDU-9 nuclear reactor. The analysis results showed that for each postulated accident the peak pressure values in the reactor headers are within the acceptance criteria given in ASME code requirements and the fuel overheating is prevented. The analysis results showed that the flow reversal through the fuel channel occurred but didn't result in any damage on the fuel bundle.