Electrochemistry of Water-Cooled Nuclear Reactors
This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.
- Research Organization:
- Pennsylvania State Univ., University Park, PA (United States)
- Sponsoring Organization:
- USDOE - Office of Nuclear Energy, Science and Technology (NE)
- DOE Contract Number:
- FG07-02ID14334
- OSTI ID:
- 890516
- Report Number(s):
- Final Report; TRN: US0703160
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
CORROSION
CRACK PROPAGATION
ELECTROCHEMISTRY
FLUID FLOW
HYDROGEN
IMPLEMENTATION
NUCLEAR POWER PLANTS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
RADIOACTIVITY TRANSPORT
REACTORS
STRESS CORROSION
THERMAL HYDRAULICS
WATER
WATER CHEMISTRY
materials science
corrosion
water-cooled nuclear reactions
electrochemistry
radiolysis