|Report date||Site||Event description|
|05000400/LER-2016-007||9 February 2017||Harris|
On October 26, 2016, the Shearon Harris Nuclear Power Plant was in a planned refueling outage. Operations was in the process of restoring the containment spray system following maintenance. During this restoration process, operations started the 'B' containment spray pump with Refueling Water Storage Tank (RWST) level below 23.4 percent. As a result, the logic to initiate containment spray switchover to the containment sump was satisfied, opening the containment sump suction valve, which established a flowpath that allowed water to be transferred from the RWST to the containment sump. Operations secured the 'B' containment spray pump and re-closed the containment sump suction valve to restore the plant to the desired configuration. During the event, the containment spray system was aligned for recirculation of the spray pump discharge back to the RWST, so no water flowed through the spray header.
The primary cause of the event was a procedural deficiency. The procedure did not establish a physical barrier to prevent the containment sump valves from opening in Modes 5, 6 and defueled. The corrective actions include revising the procedure to remove power to the containment sump valves to prevent them from opening in Modes 5, 6 and defueled.
|05000400/LER-2016-006||14 December 2016||Harris|
Between October 15 and October 19, 2016, the Shearon Harris Nuclear Power Plant (SHNPP) reactor vessel closure head penetrations were being examined. SHNPP was shut down for a scheduled refueling outage (RFO) for cycle 20 (RFO-20).
Nondestructive examinations identified four rejectable indications impacting four penetration nozzles. Indications associated with nozzles 30, 40, and 51 were indicative of primary water stress corrosion cracking (PWSCC), with the largest indication having an axial extent of 0.372 in. with a through-wall extent of 0.247 in. (39 percent). The fourth indication was identified on nozzle 23 by dye penetrant testing. This indication had a rounded profile indicative of a weld fabrication void, and was 0.307 in. on the major dimension.
The weld was fabricated during the previous outage, RFO-19. The void was originally identified during RFO-19 and was acceptable.
However, the void has since opened to unacceptable dimensions due to normal operating conditions.
A leak path assessment and a bare metal visual examination of the reactor vessel head top was completed, with no leakage identified.
The three PWSCC indications were repaired using the inside diameter temper bead weld method. The fabrication void was removed via localized grinding, with no additional welding necessary. All repairs were completed prior to exiting the refueling outage.
|05000400/LER-2016-004||7 December 2016||Harris|
On October 8, 2016, the Shearon Harris Nuclear Power Plant was reducing power to enter a planned refueling outage (RFO-20). The plant was at approximately 8 percent power in Mode 1 when the unit experienced an unplanned reactor trip with a safety injection (SI) and main steam line isolation (MSLI). A malfunction of the turbine controller during turbine mechanical overspeed trip testing caused an excessive draw of steam flow from the Steam Generators (SGs). This caused the Engineering Safety Features Actuation System instruments to detect a valid change in SG pressure and initiate a rate compensated Low Steam Line Pressure signal. This signal initiated a SI and MSLI, which in turn initiated reactor trip, turbine trip, feedwater isolation, and closed the main steam isolation valves.
Degraded equipment within the turbine controller resulted in excessive opening of the governor valves; this was caused by an inadequate supply of hydraulic oil to meet the increased system demand during testing. Insufficient hydraulic accumulator capacity was available to support system demand. One accumulator was known to be out-of-service; a second was discovered post-event. Also, a hydraulic oil pressure switch used for turbine control was not functioning properly. The equipment deficiencies have been corrected.
Changes have been made to the testing procedure to validate at least four accumulators are in service prior to testing. The Power Operation (Mode 2 to Mode 1) procedure will also be revised to validate at least four accumulators are in service. A new calibration procedure will be implemented for the deficient oil pressure switch to ensure better quality control over verifying switch function.
|05000400/LER-2016-005||7 December 2016||Harris|
On October 8, 2016, at approximately 1310 EDT, while in Mode 4 for a planned refueling outage, Shearon Harris Nuclear Power Plant experienced an undervoltage (UV) condition in the switchyard for about 1.5 seconds. This triggered the UV relays for both emergency 6.9 kV buses and for several of the non-nuclear safety 6.9kV auxiliary buses, resulting in the respective supply breakers opening. At the time of the UV, the site was experiencing high winds and rain from the effects of Hurricane Matthew. Both Emergency Diesel Generators started and loaded as designed. Operations restored offsite power at 2154 EDT after verifying stable grid behavior for an extended period. Additionally, the Containment Ventilation Isolation system and the Auxiliary Feedwater system actuated and performed as designed.
The site declared an Unusual Event at 1328 EDT for loss of offsite power to emergency buses for greater than 15 minutes. At 2049 EDT, the Unusual Event was terminated.
The causes of the UV were determined to be a line fault on the Cape Fear - West End 230 kV line and equipment deficiencies associated with the Cape Fear 230 kV Substation protection relays which prevented immediate clearing of the fault.
|05000400/LER-2016-002||20 September 2016||Harris|
On July 22, 2016, the 'A' Essential Services Chilled Water (ESCW) chiller tripped due to low oil pressure and was declared inoperable.
The cause was oil leakage from a recently replaced brass tube fitting connecting the high-pressure side of an oil pressure differential sensing line to the lubrication system. An equivalent stainless steel fitting was installed in place of the brass, and the 'A' ESCW chiller was then declared operable.
This failed brass fitting was a recently installed like-for-like replacement component following a similar 'A' ESCW chiller trip on July 15, 2016, caused by low oil pressure as a result of leakage through the original brass tube fitting.
The cause of the failed tube fitting was fatigue crack growth resulting from vibration, with stress corrosion cracking having a secondary role. The material was replaced with stainless steel, and further analysis to address the vibration is ongoing.
|05000400/LER-2016-001||1 September 2016||Harris|
On July 7, 2016, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Shearon Harris Nuclear Power Plant identified nonconforming conditions in the plant design such that specific TS equipment did not meet current design basis for protection against potential tornado missile impact. Identified systems were declared inoperable and Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance," was implemented. Compensatory measures were implemented within the time allowed by the applicable Limiting Condition(s) for Operation and the associated systems were then declared operable but nonconforming.
The two systems identified with credible impacts were the 'A' train Emergency Diesel Generator and the Main Steam Safety Relief Valves. Actions will be taken to establish compliance, either by plant modification or by employing a methodology for addressing tornado missile noncompliances.
Due to the historical nature of the issue, a specific cause for the identified vulnerabilities was not determined.
|05000400/LER-2015-004||20 August 2015||Harris||On May 4, 2015, while Harris Nuclear Plant, Unit 1 was shut down for a scheduled refueling outage in mode 5, the Operations Surveillance Test for Safety Injection, Engineered Safety Feature Response Time on Train B, was being performed. During this test the ‘A' Emergency Service Water (ESW) pump failed, resulting in a loss of flow and pressure to the ‘A' ESW header. The probable root cause of this event was determined to be a misalignment stemming from a combination of potential factors. The contributing cause of this event was determined to be that performance monitoring was not rigorous enough to identify potential pump issues prior to failure. Immediate corrective action was taken to secure the corresponding charging/safety injection pump and emergency diesel generator and to realign the Normal Service Water to the ‘A' ESW header. The ‘A' ESW pump was rebuilt with new couplings, coupling fasteners, and bearings prior to plant startup that commenced on May 13, 2015. Planned corrective actions include a procedure revision to implement a process for performing vertical alignment checks during ESW pump installation and modifications to improve ESW coupling design.|
|05000400/LER-2015-005||17 August 2015||Harris|
On June 16, 2015, while Harris Nuclear Plant (HNP) was operating at 100% in mode 1, two doors between the Reactor Auxiliary Building (RAB) and Main Steam Tunnel (MST) were opened to support a maintenance activity. These doors are credited in the high energy line break (HELB) equipment qualification and internal flooding analyses for HNP; however the opening of these doors is not addressed in these analyses. If a HELB occurred in the MST during a time in which these doors are open, the Essential Services Chilled Water system could become inoperable.
The root cause of this event was determined to be that HNP Engineering failed to develop and implement control measures for hazard barriers credited for mitigating HELB events. Immediate corrective action was taken to close the doors and issue a Standing Instruction that prohibits these doors from being blocked open during modes 1 through 4. Corrective action is planned to develop and implement an engineering change that evaluates the required passive design features needed to support HELB analysis and overall licensing bases. This engineering change will identify required passive design features and establish the necessary process to ensure barriers are appropriately identified and controlled.
|05000400/LER-2015-002||3 June 2015||Harris|
On April 4, 2015, Harris Nuclear Plant was shut down for a scheduled refueling outage in mode 5 and was performing the Remote Shutdown System Operability test. Following transfer back to the Main Control Board, the supply breakers to the normal air intake isolation dampers' motor actuators both independently tripped due to high instantaneous current from the attempted direction reversal of their respective motor actuators.
These trips caused both dampers to be in the partially open position, rendering the Control Room Envelope (CRE) boundary inoperable. The apparent cause of this event is that the HMCP model breaker/starter combination installed by a Design Change is more sensitive to peak current spikes than the original EF3 model breakers. The contributing cause associated with this event was that industry operating experience (OE) was not adequately reviewed to identify existing OE on the need to raise the trip setting on HMCP model breakers. Immediate corrective action was taken to manually close the dampers and restore integrity of the CRE boundary. The corrective action taken to address the breaker sensitivity observed was that the trip settings for the impacted HMCP model breakers installed by the Design Change were revised to add margin to the trip settings.
|05000400/LER-2015-003||28 May 2015||Harris|
On April 7, 2015, and April 9, 2015, the reactor vessel head penetration nozzles were being examined while the Harris Nuclear Plant was shut down for a scheduled refueling outage. Ultrasonic examinations identified indications that required repair in three head penetration nozzles. The indications were approximately 0.233, 0.260, and 0.297 inches long in nozzles 14, 18, and 23 respectively, and axial in orientation. The maximum through-wall extent was approximately 32%. An inspection of the exterior surfaces of the reactor head confirmed there was no leakage. The indications were repaired using the inside diameter temper bead welding process. The repairs restored compliance with the American Society of Mechanical Engineers code requirements.
The cause of the indications was attributed to primary water stress corrosion cracking. Per the requirement of 10 CFR 50.55a(g)(6)(ii)(D)(5), examinations are required to be performed on the reactor vessel head every refueling outage to identify flaws and ensure appropriate repairs are performed. This is similar to the conditions reported in Harris Licensee Event Reports 2013-001-00 and 2013-003-00.
|05000400/LER-2015-001||11 May 2015||Harris|
On March 12, 2015, while operating at 100% power in mode one, surveillance testing identified the lift setting of main steam (MS) safety valve (SV) 1MS-49 was outside of the technical specification (TS) allowed tolerance.
Testing of all other installed MSSVs identified four additional MSSVs that were also outside of the TS allowed tolerance. On April 9, 2015, while shutdown for a refueling outage, pressurizer SV 1RC-127 was found outside its TS allowable tolerance. The primary cause of the SV lift settings being outside of the TS allowed tolerance was setpoint drift incompatible with analysis specified criteria. 1RC-127 was also determined to have an operating history of being outside of tolerance when tested. The valve installed as 1RC-127 was removed from service, and all six valve lift settings have been restored to within the allowable tolerance. The 1% TS tolerance is more restrictive than the American Society of Mechanical Engineers code allowed tolerance of 3%. The primary corrective action to preclude recurrence will be implementation of a revised safety analysis that accommodates increased setpoint drift and supports revised technical specification setpoints. The condition is similar to that reported in licensee event report 2013-002-00. The small deviations beyond the allowable tolerance result in a very low safety significance.
|05000400/LER-2014-003||8 December 2014||Harris||On October 8, 2014, while operating at 100 percent power in Mode 1, Harris Nuclear Plant (HNP) personnel determined that the calculation for Diesel Fuel Oil Storage Building (DFOSB) Internal Flooding was invalid during an extent of condition evaluation. The evaluation determined that the calculation identified a single active failure to occur by the failure of a single sump pump and neglected the fact that the power source to both sump pumps was the common Motor Control Center (MCC-1-4B13). It was also determined that the internal flood event postulated for a moderate energy line break would challenge Operations' capability to respond to the event and to isolate the pipe break, prior to the flood water affecting the safety related Diesel Fuel Oil (DFO) Transfer Pumps. The DFO Transfer Pumps are used to refill the Emergency Diesel Generator (EDG) Day Tanks when the EDGs are running for an extended period of time in response to a loss of offsite power. The DFOSB fire suppression system supply was isolated promptly and a compensatory fire watch was placed at the DFOSB. Backup fire suppression is provided by staged fire extinguishers. The apparent cause was identified as a historical deficiency associated with the original design of the plant and inadequate technical rigor used to revise the calculation for DFOSB Internal Flooding in 2003.|
|05000400/LER-2014-002||11 August 2014||Harris|
On June 12, 2014, while operating at 100 percent power in Mode 1, Harris plant personnel determined that inadequate cable protection existed in control cables for the Turbine Emergency Oil Pump. The cables had no electrical protection devices (fuses) other than the circuit breaker. The control cables are routed through several fire areas, including the main control room, and under postulated conditions, could have created a common enclosure fire hazard situation. Compensatory measures (hourly fire watches) were implemented for affected areas of the plant to ensure continued public safety until electrical protection devices were installed.
The apparent cause was the historical installation of an unfused section of cable in the control circuit of the Turbine Emergency Oil Pump.
|05000400/LER-2013-004||10 April 2014||Harris|
On November 11, 2013, while in Mode 5, sample results of the 'C' waste gas decay tank (WGDT) were not properly evaluated in which hydrogen results should have been documented as greater than 4 percent and oxygen at 4.03 percent. Due to this condition, it was not recognized that the conditions of Technical Specification (TS) 126.96.36.199 were not met and required immediate suspension of all additions of waste gas to the system, reduction of oxygen to less than or equal to 4 percent by volume, then reduction of oxygen to less than or equal to 2 percent within 48 hours. HNP was in this condition for 30 calendar days until the oxygen concentration in the 'C' WGDT was reduced to below the TS limit on December 11, 2013.
The root cause was determined to be an inadequate understanding of the risks associated with degassing the waste gas system with inoperable analyzers. Corrective actions include revising procedures, training, and reinforcing expectations for procedural use and adherence. This event is considered to have no safety significance as an engineering evaluation determined that this condition did not generate any combination of explosive gases in the system.
|05000400/LER-2014-001||19 March 2014||Harris|
On January 18, 2014, while operating at 75 percent power in Mode 1, Harris Nuclear Plant manually actuated the reactor protection system to trip Unit 1 due to indications of a fire in the 480V 1D2 transformer. After the reactor trip, the auxiliary feedwater system automatically initiated and provided feedwater to the steam generators. De-energizing auxiliary bus 1D caused a temporary loss of power to the 6.9kV 1A-SA safety bus, and thus the "A" auxiliary feedwater pump which was recovered when the "A" emergency diesel generator started and re-energized the 1A-SA bus. All safety systems responded as expected during this event.
The root cause was determined to be a combination of age and various electro-magnetic conditions which over time led to failure of the 1D2 transformer. Corrective actions include transformer replacements and additional testing. The impact on safety due to this event is a small risk increase of approximately 3E-6/yr in delta core damage frequency.
|05000400/LER-2013-003||15 January 2014||Harris|
On November 18, 2013, the reactor vessel head penetrations were being examined while the Harris Nuclear Plant was shut down for a scheduled refueling outage. Ultrasonic examinations identified an indication in head penetration nozzle 37. The indication was approximately 0.46 inches long with an axial orientation. An inspection of the exterior surfaces of the reactor head confirmed there was no leakage. Nozzle 37 was repaired utilizing the inside diameter temper bead welding process, which was completed December 2, 2013. The repair restored compliance with the American Society of Mechanical Engineers code requirements.
The cause of the indication in nozzle 37 was attributed to primary water stress corrosion cracking.
Per the requirement of 10 CFR 50.55a(g)(6)(ii)(D)(5), examinations are required to be performed on the reactor vessel head every refueling outage to identify flaws and ensure appropriate repairs are performed. This is similar to the conditions reported in Harris Licensee Event Report 2013-001-00.
Shearon Harris Nuclear Power Plant, Unit 1 05000400
|05000400/LER-2013-002||19 December 2013||Harris|
On October 23, 2013, while operating at 100% power in mode 1, two main steam safety valve setpoints were found outside the +/- 1% tolerance of table 3.7-2 of technical specification 3/188.8.131.52. Three other valves were tested and setpoints were found to be within tolerance. Upon discovery of the out of tolerance conditions, the two setpoints were adjusted to within technical specification tolerances which restored compliance with the technical specifications. The cause was determined to be setpoint drift incompatible with analysis specified criteria. The corrective action will be implementation of a revised safety analysis that accommodates increased setpoint drift and supports revised technical specification setpoints.
Main steam safety valves are used to satisfy American Society of Mechanical Engineers code requirements for overpressure protection. The limiting accident analysis has more margin available than the measured deviation from setpoint, so the impact on safety is very minor.
Shearon Harris Nuclear Power Plant, Unit 1 05000400
|05000400/LER-2013-001||12 July 2013||Harris|
On May 15, 2013, while at 98% power in Mode 1, HNP commenced a Technical Specification required shut down to Mode 6 to repair a flaw that was identified in nozzle 49 of the reactor pressure vessel head.
Nozzle 49 was subsequently repaired on May 31, 2013, utilizing the inside diameter temper bead welding process. In 2012, four nozzles (5, 17, 38, 63) were identified with similar indications that exhibited characteristics of PWSCC, and were subsequently repaired using the inner diameter temper bead welding process. However, nozzle 49 was not identified as having an indication at that time.
Because the indication in nozzle 49 was identified while at power, a shut down was required by Technical Specifications.
The cause of the flaws in nozzle 49 and the other four nozzles was attributed to PWSSC. The root cause evaluation determined that the missed identification of the indication in nozzle 49 was due to the lack of mitigating programmatic governances to specify process independence and fatigue/distraction controls. The planned corrective action to prevent recurrence is to create mitigating programmatic governance for providing oversight for complex automated Non-Destructive Examination (NDE) inspections through the generation of new procedure(s).
|05000400/LER-2012-002||30 July 2012||Harris|
An investigation in May 2012 revealed that surveillance tests for containment penetration overcurrent protection devices had not been properly scheduled. Testing was performed on the components with missed surveillances over the next several weeks. On May 31, 2012, the breaker for Pressurizer Heater Bank C could not be tested due to a broken handle, which precluded demonstration of acceptable performance. Subsequently, on June 2, the B Reactor Coolant Pump overcurrent protection timing relay did not meet acceptance criteria for its surveillance test. Missed surveillances are required to be reported if subsequent testing does not demonstrate acceptable results.
Root and Contributing Causes to the missed surveillances are historical in nature resulting from activities in the 1980s and 1990s. An extent of condition evaluation resulted in the testing of 30 breakers and six timing relays, with all but the subject two components described above passing the subsequent surveillance testing. The two components that did not pass subsequent testing were replaced and tested satisfactorily.
Completed corrective actions include replacement and satisfactory testing of the breaker and relay.
Planned corrective actions include revising impacted plant procedures and design documents.
|05000400/LER-2012-001||20 June 2012||Harris||At 5:00 AM on April 21, 2012, Harris Nuclear Plant (HNP) was in a refueling outage, in Mode 4 (Hot Shutdown), cooling down to cold shutdown. During a surveillance test, two Main Steam Isolation Valves (MSIVs) did not close with acceptable stroke times. The two valves were declared INOPERABLE resulting in entry into Technical Specification (TS) 3.0.3. The plant continued to cool down as planned, entered Mode 5 (Cold Shutdown) at 11:27 AM, and exited TS 3.0.3. Troubleshooting revealed high friction within all three MSIVs with the root cause being unexpected long-term corrosion of the valve piston rings. Primary contributing causes were that the MSIVs were not properly categorized in the Air Operated Valve Program, and opportunities to identify potential valve degradation were not recognized. Corrective actions to prevent recurrence include replacing the valve piston rings with upgraded rings made of a material less susceptible to corrosion and implementing a diagnostic testing program on the MSIVs.|