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On August 15, 2017, Callaway Plant was in Mode 1 at 100 percent power. During evaluation of protection for safety-related equipment from the damaging effects of tornados, Callaway Plant personnel determined that the minimum-flow recirculation lines for the turbine-driven auxiliary feedwater pump (TDAFP) and both motor-driven auxiliary feedwater pumps (MDAFPs) could be damaged if a postulated tornado-generated missile were to penetrate the condensate storage tank (CST) valve house and strike the lines. In response, Operations declared all three auxiliary feedwater pumps inoperable.

Compensatory measures were implemented consistent with Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance." Upon completion of the initial compensatory measures, the TDAFP and MDAFPs were declared Operable but nonconforming.

Subsequent to the condition identified on August 15, 2017, continued investigation of tornado missile vulnerabilities led to discovery that the exposed steam exhaust stacks for the main steam safety valves and atmospheric steam dump valves, as well as the exposed vents for the diesel generator fuel oil storage and day tanks, were also susceptible to tornado missile damage to the extent that compliance with General Design Criterion 2 is not ensured. Compensatory measures were then promptly implemented for these conditions, as well, in accordance with EGM 15-002 such that the affected systems have been evaluated to be nonconforming but Operable.

It has been determined that the identified noncomformances are an original plant design legacy issue. Long-term resolution for establishing compliance is under development and will be completed within the time frame described in the EGM.

On September 13, 2016, and September 14, 2016, during plant walk downs by Engineering and the NRC Senior Resident inspector, pressurized fire protection piping in the Service Water Intake Structure was found to not be shielded against a Moderate Energy Line Break (MELB), resulting in inoperability of one train of Service Water for both units.

During extent of condition walk downs conducted on October 6, 2016, October 10, 2016, November 17, 2016, December 5, 2016, and December 22, 2016, additional piping in the Unit 1 and Unit 2 Safeguards and Auxiliary Buildings was found to not be shielded against a MELB, resulting in inoperability of one train of various.safety related equipment for both units. The most likely cause of this event was the methodology used to conduct the initial MELB walk downs was flawed and allowed some MELB threats to be missed.

Corrective actions include shielding the affected piping, performing a 100 percent walk down of rooms containing MELB piping identified for shielding, and revising the systems interaction program maintenance procedure. I All times in this report are approximate and Central Time unless noted otherwise.

, On April 19, 2016, at approximately 1450 hours, it was discovered that a maintenance activity performed between 2010 hours on August 26, 2014 and 0143 hours on August 27, 2014, had rendered the Startup Transformer (X4) and the standby Emergency Diesel Generators (EDG) (X-107A&B) unable to automatically supply power to Buses A5 and A6, due to the breaker interlock that would prevent Startup Transformer breakers (152-504 and 152-604) and standby EDG breakers (152-509 and 152-609) from closing, when Bus A8 to Bus A5 breaker (152-501) and Bus A8 to Bus A6 breaker (52-601) are in the TEST position and CLOSED. During the maintenance activity, the plant was operating at 100 percent power and the_ Unit Auxiliary Transformer (X3) was providing power to Emergency Buses A5/A6.

The functional testing of negative sequence relays (146-600/A and B) and 23kV feed undervoltage relays (127-600A/1 and 2, and 127-600B/1 and 2) created a test configuration, lasting less than 1-hour, whereby power to Buses A5 and A6 was not automatically available from either the startup transformer or from the EDGs. As a result, Limiting Conditions for Operation (LCO) Action Statement 3.9.6.2 was not met.

The root cause is that the decision to perform the described surveillance testing online, instead of during cold shutdown, lacked sufficient rigor to ensure compliance with Technical Spepifications. Corrective actions will establish and institutionalize expectations and accountability for station leadership regarding consequence-biased decision-making and effective risk manaaement. There was no impact to_public health and safety.

On April 24, 2014, during a review of previous conditions affecting equipment qualification it was identified that the environmental qualification of Namco EA180 series limit switches were not being properly maintained per vendor requirements. This condition was not verbally reported at the time of discovery as the condition was identified and resolved while the plant was in an extended shutdown.

A cause evaluation was completed and determined that technical requirements from the vendor manual for maintaining environmental qualification of the Namco EA180 series limit switches were not captured in the applicable plant procedure.

The applicable plant procedure has been revised to include vendor information for maintaining environmental qualification of the limit switches. The limit switch top cover gasket and screw assemblies for all environmentally qualified Namco EA180 series limit switches were replaced and torqued in accordance with vendor requirements.

On October 22, 2013 it was discovered that unfused ammeter indication circuits associated with the Nine Mile Point Unit 1 (NMP1) safety-related Direct Current (DC) buses could short circuit due to a fire in the circuit cable routing. This ground fault equivalent hot short could cause the cable to self-heat and lead to secondary fires. The unanalyzed secondary fires could adversely affect safe shutdown equipment and potentially cause the loss of the ability to safely shutdown as required by 10 CFR 50 Appendix R.

This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(B) as a condition that resulted in the nuclear plant being in an unanalyzed condition that significantly degraded plant safety, and 10 CFR 50.73(a)(2)(ix)(A) as a condition that as a result of a single cause could have prevented the fulfillment of a safety function for two or more trains or channels in different systems.

The cause of this event is that the equipment design issue was not recognized as an unacceptable configuration. The design issue is associated with an evolving industry understanding of the 10 CFR 50 Appendix R common enclosure scope.

Corrective actions include the isolation of the affected circuits at the DC bus in order to prevent the condition from occurring, and the development, issuance and installation of a plant modification to install fuses on the safety-related DC ammeters at NMP I to ensure adequate circuit protection to prevent the propagation of fires in additional areas due to overcurrent conditions.

Between October 2009 and March 2012, Comanche Peak Nuclear Power Plant (CPNPP) Units 1 and 2 experienced three failures of safety related breaker control devices that caused Technical Specification equipment to become inoperable. Two of the failures were on Unit 1 and one of the failures was on Unit 2. On June 14, 2012, a failure analysis concluded that these failures were caused by a material weakness of the control devices' contact carrier frames due to a manufacturing defect. Less than adequate control of the manufacturing process increased the susceptibility to shock loading of the cured phenolic material used to make the contact carrier frames. On June 20, 2012, CPNPP determined that this condition was reportable per 10CFR50.73(a)(2)(ix)(A) and 10CFR21. Corrective actions include replacement of the potentially affected safety related breaker control devices, development of receipt inspection testing criteria to be used for breaker control devices, issuing an Operating Experience Report on this event, and notification of the breaker vendor (ABB Inc.).

All times in this report are approximate and Central Time unless noted otherwise.

On December 14, 2009, at 0800 hours, Comanche Peak Nuclear Power Plant (CPNPP) Unit 2 was in Mode 1 at 100% power. A channel calibration on Unit 2 Train B Neutron Flux Monitoring System (Gamma-Metrics) resulted in the inoperability of channels in different systems: Unit 2 containment pressure channel 4, and Channel 4 of the Overtemperature N-16 and Overpower N-16 reactor trips.

When the test equipment being used for the channel calibration was disconnected from the Gamma- Metrics system, the affected control room indications returned to normal. The cause of this event was grounding the Gamma-Metrics system to current paths and ground loops which served to couple a voltage potential through the plant grounding system to the affected plant parameters. Completed corrective actions include: 1) A Maintenance Standing Order was issued to control the use of test equipment that has the ability to ground a floating circuit, and (2) The shield cable ground was corrected for the Unit 2 containment pressure channel 3 and 4 main control board indicators. Planned corrective actions include procedure changes to improve the control and use of test equipment.

All times in this report are approximate and Central Standard Time unless noted otherwise.

On August 16, 2002, during the development of the new AOV evaluation and assessment program at Point Beach Nuclear Plant (PBNP), NMC Engineers determined that 1MS-5959, one of two steam generator (SG) blowdown isolation valves on Unit 1, may have been unable to shut with full differential pressure of 1085 psig across the valve.

The new AOV evaluation and assessment program is used to determine whether the valve's performance will meet design basis functions under prescribed operating conditions. In this case, we evaluated that 1MS-5959 would not have shut under those conditions. Since this is a new evaluation method, it was not used when the valve was diagnostically tested in 1999.

This condition was documented in the PBNP corrective action program as CAP029065. Interim action was taken to place the valve in the required shut position to ensure that minimum feed water flow requirements would be maintained in the event of a loss of normal feed water accident. Following diagnostic testing, adjustment of the spring closing force of this AOV, and successful completion of post maintenance testing, 1MS-5959 was restored to full operability, and Unit 1 "B" SG blowdown was restored. The remaining Unit 1 and 2 SG blowdown isolation valves at PBNP were evaluated and determined to be capable of shutting against the maximum differential pressure assumed within the calculation.

This event is reportable as an unanalyzed condition that had the potential to significantly degrade plant safety. The safety assessment concluded that the overall safety significance of this event was low.