Corrosion Mechanisms of Ni-Base Alloys in Pressurized Water Reactor Primary Conditions

Oxidation mechanism of Alloy 690 has been investigated in Pressurised Water Reactor (PWR) primary coolant conditions (325°C, aqueous hydrogenated media). Experiments performed with gold marker and RBS technique reveal that the passive film formation is the consequence of an anionic mechanism. This result is confirmed by experiments achieved with two sequences of corrosion in a H2 16O media and in a mixed H2 16O/ H2 18O media. The localisation of 18O by SIMS analysis in the thin passive layer underlines an oxidation mechanism due to oxygen diffusion by short circuits (like grain boundaries) in the oxide scale. Moreover grain boundary diffusion coefficient in chromite like oxide was estimated to be in the range 2 10-18 – 1 10-17 cm2.s-1 and compared to values extrapolated from higher temperature.

Download Full-text

Development of a Thermal Fatigue Screening and Evaluation Methodology for Pressurized Water Reactor Plants

Elevated Temperature Design and Analysis, Nonlinear Analysis, and Plastic Components ◽

10.1115/pvp2004-2995 ◽

2004 ◽

Cited By ~ 2

Author(s):

J. D. Keller ◽

A. J. Bilanin ◽

S. T. Rosinski

Keyword(s):

Thermal Cycling ◽

Development Program ◽

Fatigue Cracking ◽

Evaluation Methodology ◽

Experimental Investigations ◽

Pressurized Water Reactor ◽

Primary Coolant ◽

Modeling Tools ◽

Pressurized Water ◽

Water Reactor

Thermal cycling has been identified as a mechanism that can potentially lead to fatigue cracking in un-isolable branch lines attached to pressurized water reactor (PWR) primary coolant piping. A significant research and development program has been undertaken to understand the mechanisms causing thermal cycling and to develop models for predicting the thermal-hydraulic boundary conditions for use in piping structural and fatigue analysis. A combination of first-principles engineering modeling and scaled experimental investigations has been used to formulate improved thermal cycling modeling tools. This paper will provide an overview of the model development program, a summary of the supporting test program, and a description of the thermal cycling model structure. Benchmarking of the thermal cycling model against several PWR plant configurations is presented, demonstrating favorable comparison with cases where thermal stratification and cycling has been previously observed.

Download Full-text

A high-resolution characterization of the initiation of stress corrosion crack in Alloy 690 in simulated pressurized water reactor primary water

Corrosion Science ◽

10.1016/j.corsci.2019.108243 ◽

2020 ◽

Vol 163 ◽

pp. 108243 ◽

Cited By ~ 2

Author(s):

Wenjun Kuang ◽

Gary S. Was

Keyword(s):

High Resolution ◽

Stress Corrosion ◽

Stress Corrosion Crack ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Alloy 690 ◽

Primary Water

Download Full-text

Calculation of maximum equilibrium pressure and temperature inside the containment of a pressurized water reactor following rupture of the primary coolant circuit

Nuclear Engineering and Design ◽

10.1016/0029-5493(66)90058-6 ◽

1966 ◽

Vol 4 (3) ◽

pp. 331-338

Author(s):

R.B. Galli ◽

B.G. Lord

Keyword(s):

Pressurized Water Reactor ◽

Coolant Circuit ◽

Primary Coolant ◽

Pressurized Water ◽

Equilibrium Pressure ◽

Water Reactor

Download Full-text

Linking Grain Boundary Microstructure to Stress Corrosion Cracking of Cold-Rolled Alloy 690 in Pressurized Water Reactor Primary Water

CORROSION ◽

10.5006/0808 ◽

2013 ◽

Vol 69 (10) ◽

pp. 953-963 ◽

Cited By ~ 40

Author(s):

S.M. Bruemmer ◽

M.J. Olszta ◽

M.B. Toloczko ◽

L.E. Thomas

Keyword(s):

Grain Boundary ◽

Stress Corrosion ◽

Stress Corrosion Cracking ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Alloy 690 ◽

Primary Water ◽

Cold Rolled ◽

Boundary Microstructure

Download Full-text

Concentration of Radiocarbon and Its Chemical Forms in Gaseous Effluents, Environmental Air, Nuclear Waste and Primary Water of a Pressurized Water Reactor Power Plant in Hungary

Radiocarbon ◽

10.1017/s0033822200030976 ◽

1995 ◽

Vol 37 (2) ◽

pp. 497-504 ◽

Cited By ~ 23

Author(s):

Mihály Veres ◽

Ede Hertelendi ◽

György Uchrin ◽

Eszter Csaba ◽

István Barnabás ◽

...

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Discharge Rate ◽

Chemical Forms ◽

Pressurized Water Reactor ◽

Primary Coolant ◽

Pressurized Water ◽

Water Reactor ◽

Environmental Air

We measured airborne releases of 14C from the Paks Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). Two continuous stack samplers collect 14C in 14CO2 and 14CnHm chemical forms. 14C activities were measured using two techniques; environmental air samples of lower activities were analyzed by proportional counting, stack samples were measured by liquid scintillation counting. 14C concentration of air in the stack varies between 80 and 200 Bqm−3. The average normalized yearly discharge rates for 1988–1993 were 0.74 TBqGW−1ey−1 for hydrocarbons and 0.06 TBqGW−1ey−1 for CO2. The discharge rate from Paks Nuclear Power Plant is about four times higher than the mean discharge value of a typical Western European PWR NPP. The higher 14C production may be apportioned to the higher level of nitrogen impurities in the primary coolant. Monitoring the long-term average excess from the NPP gave D14C = 3.5‰ for CO2 and D14C = 20‰ for hydrocarbons. We determined 14C activity concentration in the primary coolant to be ca. 4 kBq liter−1. The 14C activity concentrations of spent mixed bed ion exchange resins vary between 1.2 and 5.3 MBqkg−1 dry weight.

Download Full-text