10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis for those structures, systems, and components to which the appendix applies are correctly translated into specifications. Contrary to Criterion III, Exelon failed to correctly translate the design basis for 15
PCIVs on Unit 1 and 15
PCIVs on Unit 2 into specification for the motor operators for the valves. This resulted in the valves not being able to perform their intended safety function under certain conditions following a loss of coolant accident with offsite power remaining available. Exelon entered this issue into the
CAP as
IR 1402693 and 1416070. The inspectors determined that the finding was of very low safety significance (Green) in accordance with Section B of Exhibit 3 of NRC
IMC 0609, Appendix A, The
Significance Determination Process for Findings At-Power, because it did not represent an actual open pathway in the physical integrity of reactor containment. In addition, all of the effected penetrations, with the exception of two, had another
primary containment isolation valve that was not impacted by the design issue. Conservatively processing the two penetrations that did not contain a non-effected valve through NRC
IMC 0609, Appendix H, Containment Integrity
Significance Determination Process, determined that they were not risk significant from a
large early release frequency standpoint. One penetration (suppression pool clean-up suction line) did not connect with the drywell atmosphere or
reactor coolant system. The other penetration (
reactor water cleanup suction line) is a closed system and the design error would not have affected the ability of the systems
primary containment isolation valves to isolate following their design basis event (intersystem loss of coolant accident outside of containment).
