Apparatus and method for limiting and analyzing stress corrosion cracking in pressurized water reactors

Abstract

A method to assess and predict pressurized water stress corrosion cracking in operational nuclear power plants and the effect of adding zinc compounds into a reactor coolant system of the nuclear power plant.

1 . A method to evaluate an effect of applying a zinc compound to a reactor coolant system of a pressurized water reactor, comprising: 
 quantitatively assessing a pressurized water reactor stress corrosion cracking initiation rate of a candidate system through analysis of operational eddy current testing data and pressurized water stress corrosion cracking failure history using empirical relationships;    determining an extent of damage to the candidate system;    approximating when zinc addition to the system will mitigate pressurized water stress corrosion cracking;    quantitatively assessing pressurized water stress corrosion cracking initiation benefit for high-concentration (≧10 ppb) and low-concentration (<10 ppb) zinc addition programs; and    calculating a pressurized water reactor stress corrosion cracking initiation benefit from zinc addition at low concentrations (<10 ppb) in the reactor coolant system.    
   
   
       2 . The method to evaluate the effect of applying a zinc compound to a reactor coolant system of a pressurized water reactor according to  claim 1 , further comprising: 
 applying zinc acetate to the reactor coolant system at concentrations of approximately between 1 to 10 parts per billion for pressurized water stress corrosion cracking mitigation.    
   
   
       3 . A method of controlling pressurized water reactor stress corrosion cracking, comprising: 
 providing a zinc injection system;    determining a need for mitigation of pressurized water reactor stress corrosion cracking; and    injecting zinc into the pressurized water reactor coolant, above the solubility of zinc chromate and below 10 parts per billion, the injection of zinc reducing an incidence of pressurized water reactor stress corrosion cracking.    
   
   
       4 . The method according to  claim 1 , wherein the step of quantitatively assessing a pressurized water reactor stress corrosion cracking initiation rate of a candidate system through analysis of operational eddy current testing data and pressurized water stress corrosion cracking failure history, using empirical relationships compress calculating a normalized degradation rate defined as a number of tubes with new pressurized water stress corrosion cracking indications divided by a number of rotating coil examinations in a region of the number of tubes.  
   
   
       5 . The method according to  claim 4 , further comprising: 
 applying a temperature scaling factor adjustment to the data to normalize differences between plant locations and between different plants.    
   
   
       6 . The method according to  claim 4 , further comprising: 
 developing a database of degradation rates, the database adjusted to a common reference temperature.    
   
   
       7 . The method according to  claim 6 , wherein the database is adjusted to the effective full power years of plant operation.  
   
   
       8 . The method according to  claim 5 , further comprising: 
 developing a probabilistic predictive tool to trend and predict degradation in the nuclear plant, the probabilistic predictive tool correlating effective full power years of operation and the normalized degradation rate.    
   
   
       9 . The method according to  claim 8 , wherein the probabilistic predictive tool is developed from a change in a Weibull slope of PWSCC initiation plotted data before and after zinc addition.  
   
   
       10 . The method according to  claim 1 , further comprising: 
 calculating a magnitude of PWSCC mitigation due to zinc injection at nuclear power plants as a function of a mass of zinc incorporated into surface oxides of the reactor coolant system.