Evaluation of zinc addition on fuel cladding corrosion at the Halden test reactor. Final report
Experimental studies have shown that addition of zinc to a PWR environment reduces the general corrosion rates of materials in the primary system and delays the initiation of primary water stress corrosion cracking (PWSCC) in Alloy 600. In order to provide an early warning of any potential adverse effects on the fuel cladding, corrosion studies were initiated at the Halden test reactor. These tests were carried out in a PWR rig inserted in the Halden reactor core. The rig simulated thermal hydraulic and coolant conditions typical of a MR. It had two flow channels where the fuel rod segments were exposed to the coolant under irradiation flux. Selected pre-characterized rodlets with fresh and pre-irradiated standard and low-tin Zircaloy-4 material were irradiated for three cycles. First cycle lasted for 110 effective full power days (EFPDs), the second for 95 EFPDs and the last 62 EFPDs. The cladding corrosion behavior was monitored by initial, interim and final oxide thickness measurements by eddy current lift-off probe. Crud sampling was performed in both channels after cycle 1 and 2. Destructive post-irradiation examinations (PIE) of two rodlets, irradiated during cycle 1 and 2, have also been completed at the conclusion of the in-pile testing. This report presents the results on oxide thickness measurements, irradiation history and water chemistry data, and the PIE.
- Research Organization:
- Electric Power Research Inst. (EPRI), Palo Alto, CA (United States); Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt
- Sponsoring Organization:
- Electric Power Research Inst., Palo Alto, CA (United States)
- OSTI ID:
- 399734
- Report Number(s):
- EPRI-TR-106357; TRN: 96:006378
- Resource Relation:
- Other Information: PBD: Aug 1996
- Country of Publication:
- United States
- Language:
- English
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