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During an investigation of a nitrogen leak inside the Unit 2 containment, Nitrogen Accumulator Relief Valve (RV) RV-355 was found to be leaking. The leak caused the pressure in the back up nitrogen accumulator supply to Power Operated Relief Valve (PORV) PCV-455C to decrease to a level that made the PORV inoperable. Based on a review of the ti-end data for nitrogen usage in the containment, it is conservatively assumed that RV-355 had been degraded since December 1, 2016, rendering the PORV inoperable for longer than permitted by Technical Specifications.

The presumptive cause was inadequate instructions provided in plant procedures for placing a new nitrogen bottle in service. These instructions did not provide a sequence that assures system pressure transients are mitigated. This may have caused excessive pressure excursions resulting in multiple lifts of RV-355 which resulted in damage to the RV 0-ring seat and a nitrogen leak path.

Corrective actions include replacing RV-355 and revising procedures to provide instructions on placing nitrogen supply bottles in service to maintain back pressure and minimize pressure transients on the nitrogen system.

This event did not affect the health and safety of the public.

On August 13, 2014, while performing scheduled maintenance on Unit 2 Emergency Diesel Generator (EDG) 2-2, Diablo Canyon Power Plant (DCPP) identified a failed Inlet-to-Fuel-Header capscrew on Engine Cylinder 1L. As part of subsequent inspections to determine whether a similar condition existed on any of the other Unit 1 or Unit 2 EDGs, a degraded capscrew was identified on EDG 2-3 Cylinder 8L.

No capscrew issues were identified on the Unit 1 EDGs or on Unit 2 EDG 2-1. EDG 2-3 was declared inoperable at 1631 on August 14, 2014, resulting in two of three EDGs being inoperable at the same time, which requires ensuring at least two EDGs are operable within 2 hours, or be in Mode 3 within the following 6 hours.

Although the capscrew on EDG 2-3 was successfully replaced within 2 hours, during fuel system fill and vent following corrective maintenance, a fuel oil leak from the belt driven fuel oil booster pump occurred. Because repairs of EDG 2-3 could not be completed within the time permitted by Technical Specification 3.8.1 for two EDGs inoperable, a Unit 2 plant shutdown commenced. On August 14, 2014, at 2351 hours, Unit 2 entered Mode 3.

The cause of the failed capscrew on EDG 2-2, and the degraded capscrew on EDG 2-3, was determined to be high cycle fatigue. The cause of the fuel oil booster pump leak was determined to be a manufacturing defect combined with high seal annulus pressure during fuel oil system priming. Corrective actions include replacement of all capscrews with an improved material design, and incorporation of updated vendor guidance and updated fuel system priming instructions into station procedures.

This event did not adversely affect the health or safety of the public.

On June 8, 2013, at 08:40 PDT, with Diablo Canyon Power Plant (DCPP) Unit 2 in Mode 1 at 100 percent power, Emergency Diesel Generator (EDG) 2-3 failed to complete a scheduled surveillance run. Cyclic fatigue failure of a wire lug in the EDG 2-3 current differential protection circuit caused an automatic EDG shut down 21 hours and 42 minutes into the 24-hour load run. DCPP determined the last time EDG 2-3 would have been able to complete its 24-hour surveillance run was greater than the technical specification allowed outage time. Additionally, during the time that EDG 2-3 was unable to complete its load run, EDGs 2-1 and 2-2 also had been declared inoperable on several occasions.

DCPP determined that a vibrating terminal block cover induced cyclical fatigue in the wire lug, causing it to fail.

DCPP replaced the broken wire lug and permanently removed the cover. Additionally, DCPP will revise a procedure to periodically inspect for wear on the wires and lugs.

This condition did not adversely affect the health and safety of the public.

Refueling Outage Cycle 17, Unit 2 source range (SR) instrument N-32 experienced an unexpected increase in indicated counts per second (cps). Other available SR indications showed no rise in cps. At the time, SR instrument N-31 was inoperable. Since operators considered N-32 inoperable while N-31 was already inoperable, the audible count rate in the control room was no longer reliable. DCPP determined this condition constituted a loss of a safety function required to maintain the reactor in a safe shutdown condition and was reportable in accordance with 10 CFR 50.73(a)(2)(v)(A).

Following a vendor failure analysis, DCPP determined that a discontinuity in the cable insulation shield caused the N-32 high count rate readings. DCPP replaced the faulted cable. This condition did not adversely affect the health and safety of the public.

On October 11, 2012, at 12:08 PDT, the Diablo Canyon Power Plant (DCPP) Unit 2 500kV line differential relay actuated, resulting in a unit trip. The 500kV coupling capacitor voltage transformer (CCVT) bushing experienced a flashover to ground, resulting in a unit trip and turbine trip. With the turbine tripped and Unit 2 operating above the 50 percent power permissive, the reactor protection system initiated a reactor trip as designed. All plant equipment, including the auto-start of the auxiliary feedwater (AFW) system, responded as designed.

DCPP determined that the bushing failed because the insulator minimum creepage distance was not consistent with industry codes and standards for its operating environment. When the bushing was replaced in 2011, DCPP staff made unvalidated assumptions, and over-relied on industry experts. DCPP will move CCVT metering to an area with lower contamination levels, and reinforce expectations to review current industry codes and standards.

Additionally, an unintended AFW pump restart occurred following this event as a result of a human performance error that resulted in a procedure not being revised following a plant modification. DCPP will revise procedure supporting documents and train procedure writing staff on use of the supporting documents to identify all changes required by a plant modification.

On March 26, 2011 at 1449 PDT operators at Diablo Canyon Power Plant (DCPP) Unit 2 manually initiated a reactor trip in response to loss of main feedwater pump 2-1. The main feedwater pump 2-1 tripped after non-radioactive water sprayed on its control console. The water spray was caused by leakage from the flange of the relief valve on the feedwater heater 2-1A condenser dump valve line. All rods fully inserted on the reactor trip. The reactor was cooled by the auxiliary feedwater system with the condenser in service. All systems performed as designed with no unexpected pressure or level transients.

Automatic main feedwater isolation, auxiliary feedwater actuation, and steam generator blowdown isolation occurred as expected.

ENS notification 46701 provided the 4-hour non-emergency event report for the manual reactor trip of DCPP Unit 2 in accordance with 10 CFR 50.72(b)(2)(iv)(B), RPS Actuation (scram). Additionally, this notification provided the 8-hour non-emergency event report of the automatic actuation of the auxiliary feedwater system as a result of the reactor trip in accordance with 10 CFR 50.72(b)(3)(iv)(A), Specified System Actuation.

The cause of the event was determined to be a failed gasket for the relief valve due to bolt torquing practices. Corrective actions include revising DCPP's bolt torquing procedure to provide bolt torquing specifications for bolted connections.

As part of an apparent cause evaluation, Pacific Gas and Electric Company (PG&E) identified an incorrect insulation configuration, installed in 2010, on the thermal extension piping that houses the wires for the wide range (WR) reactor coolant system (RCS) resistance temperature detectors (RTDs). The insulation configuration, as installed, trapped heat inside the thermal extension piping and overheated the wires. The cause of the incorrect configuration of the insulation was insufficient guidance in the associated work package instructions.

An engineering analysis completed on October 5, 2015, determined that the eight WR RCS RTDs had either failed or were operating outside the environmental qualification temperature range. As a result, on October 5, 2015, PG&E determined that the required number of WR RTDs would not have been operable and therefore a violation of Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," had occurred.

As part of the corrective actions, PG&E replaced all eight WR RTDs, restored the insulation per the design requirements, revised the drawings for Unit 1 WR RTDs to provide adequate level of detail, and revised the work order to include the correct drawing and level of details for proper installation of all WR RTDs. This event did not adversely affect the health or safety of the public.

On December 31, 2014, while performing a walkdown as part of a surveillance test procedure, plant personnel identified through-wall seepage in a Diablo Canyon Power Plant Unit 1 socket weld inside containment that provides a flow path to a relief valve protecting a common portion of both trains of the residual heat removal system. Subsequent cleanup of the boric acid accumulation revealed active seepage of 30 drops per minute. A visual inspection identified that the source of the seepage was a circumferential crack on the socket weld.

Pacific Gas and Electric Company determined that the root cause of the cracked socket weld was containment fan cooler unit (CFCU) vibration inducing a resonant condition in the residual heat removal piping that generated stresses above the material endurance limit of the socket weld. Corrective actions included replacing two socket welds, modifying pipe supports, and correcting the condition causing the CFCU vibrations.

This condition did not have an adverse effect on the health and safety of the public.

On March 9, 2010, while analyzing the consequences of a postulated sustained degraded grid voltage, Pacific Gas and Electric (PG&E) concluded both Units 1 and 2 were in an unanalyzed condition. On March 9, 2010, at 23:39 (EST), PG&E reported this unanalyzed condition to the NRC in accordance with 50.72(b)(3)(ii)(B) (reference NRC Event Notification Number 45754).

The postulated sustained degraded voltage condition could have resulted in multiple safety-related pump motors, tripping overcurrent relays. As a result, these pumps would not be immediately available to mitigate a postulated accident and is considered a safety system functional failure.

PG&E had not considered this postulated worst case degraded voltage condition credible, and had therefore not analyzed for it. As a result, technical specification (TS) surveillance requirement (SR) 3.3.5.3 values are nonconservative, and several TS Actions were not met.

The immediate compensatory measure was to raise the first level undervoltage relay setpoints on the vital buses of both Units 1 and 2, thus transferring loads to emergency diesel generators prior to tripping motors on overcurrent. PG&E will submit a license amendment request to establish conservative TS SR 3.3.5.3 undervoltage relay settings.

The apparent cause was misinterpretation of the undervoltage relay design criteria in the 1970's.