An NRC identified noncited violation of
10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for Energy Northwests failure to adequately review a design change to the facility in 1994. The design change installed a bypass line around a
residual heat removal pump
shutdown cooling suction
header isolation valve to bleed pressure from the
header. This would be done in the event of leakage past the
shutdown cooling suction
header reactor coolant system pressure isolation valves. The design change failed to consider the thermal effects of introducing hot
reactor coolant system water into the
residual heat removal shutdown cooling suction
header at a design maximum flowrate of 0.75 gpm. As a result, operation of the bypass line would have resulted in saturation conditions being achieved in the suction
header causing flashing across the isolation valves and potentially degrading the valve disk and seating surfaces. This could result in increased
reactor coolant system leakage past the isolation valves beyond the capacity of the bypass line. However, in the event of leakage in excess of the ability of the bypass line, Energy Northwest would have received a control room alarm which would have alerted operators to the degraded condition allowing the operators to take prompt action to define the actual leakage and to take actions as needed. Energy Northwest entered the issue into the corrective action program and took immediate action to monitor suction
header temperature with the bypass line in service to assure that saturation conditions would not develop. This finding was more than minor because it was a design control issue which affected the
initiating events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, degradation of
reactor coolant system pressure isolation valves would occur with the bypass line in service at the maximum allowable design flowrate. This was considered to be a primary system loss of coolant accident initiator contributor (i.e. intersystem loss of coolant accident). The finding was determined to be of very low risk significance (Green) because assuming worst case degradation, the finding would not result in exceeding any Technical Specification limits for
reactor coolant system leakage. Additionally, the finding would not have likely affected other mitigation systems resulting in a total loss of their safety function. A crosscutting aspect was not identified due to the performance deficiency occurring in 199
