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 Report dateSiteEvent description
05000272/LER-2017-0018 January 2018Salem

On November 9, 2017 at approximately 2300, Salem Unit 1 was operating in MODE 3 when operators found steam leaking into the mechanical penetration area outside containment. Operators entered S1.0P-AB.STM-0001, Excessive Steam Flow, and dispatched operators to locate and isolate the leak.

Operators determined the steam was from the 14 steam generator through normally closed valves 14GB47 and 14GB48 steam generator blowdown (WI) line nitrogen supply valves. The steam leak was isolated at 2314 when operators closed normally open manual valve 14GB3.

This report is made per 10CFR50.73(a)(2)(i)(B), Any operation or condition which was prohibited by the plants Technical Specifications.

This was caused by human performance. Procedures will be revised to assure containment integrity exceptions are tracked and open valves are closed while sampling during the sparging process.

NRC FORM 3116 (04-20171

05000272/LER-2015-00216 November 2016Salem

On August 5, 2014, control room operators identified one steam generator (SG) protection level channe indicator as drifting high and approaching its 3 percent level deviation limit. Subsequent troubleshooting performed on October 10, 2014 identified the level transmitter as inoperable due to exceeding its Technical Specification (TS) calibration acceptance criteria. A past functionality evaluation was completed on February 12, 2015. This evaluation determined that the best estimate for when the transmitter would have exceeded its TS calibration acceptance criteria was August 19, 2013.

The cause of the SG protection level channel drifting was due to premature failure of its level transmitter due to a manufacturing defect. Vendor testing confirmed that a center diaphragm oil leak was the cause of the transmitter output drift.

This report is made in accordance with 10CFR50.73(a)(2)(i)(B), Any operation or condition which was prohibited by the plant's Technical Specifications.

05000311/LER-2016-00631 October 2016Salem

At approximately 1500 hours on August 31, 2016, the 21 Reactor Coolant Pump (RCP) tripped resulting in an automatic reactor trip on low flow in one reactor coolant loop above the P-8 permissive (36% power permissive). As expected, the 21, 22 and 23 Auxiliary Feedwater (AFW) pumps started on low steam generator level following the unit trip. Unit 2 was stabilized in Mode 3 at normal operating temperature and pressure with the 22, 23 and 24 RCPs in-service.

The trip of the 21 RCP was caused by a Service Water (SW) leak that developed on the 22 Containment Fan Coil Unit (CFCU) motor cooler. The SW leaked on to the 21 RCP motor lead containment penetration and the motor leads in the termination box. This caused the A and C phase instantaneous overload relays to actuate causing the trip of the RCP and the subsequent reactor trip on low flow in one reactor coolant loop.

This report is being made in accordance with 10CFR50.73 (a)(2)(iv)(A), "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B)," for this event actuation of the Reactor Protection System (RPS) and the Auxiliary Feedwater System (AF).

05000311/LER-2016-00529 August 2016Salem

On 6/28/16 at 04:22 Salem Unit 2 automatically tripped from 100% power on Generator Protection. The reactor trip was initiated due to a Main Turbine (MT) trip caused by a Main Generator ProtectiOn signal.

All emergency core cooling systems and emergency safeguards feature systems functioned as expected.

The motor driven and steam driven auxiliary feedwater pumps started as expected on steam generator low level. Operators stabilized the plant in Mode 3 with decay heat removal via the main steam dump valves and auxiliary feedwater system.

Investigation identified that a broken current transformer core ground wire internal to the A Main Power Transformer (MPT) was intermittently touching the X1 and X2 low voltage connections inside the transformer bushing compartment causing a ground fault. This caused the turbine generator trip.

This report is being made in accordance with 10CFR50.73 (a)(2)(iv)(A), "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B)," for this event actuation of the Reactor Protection System (RPS) and the Auxiliary Feedwater System (AF).

05000272/LER-2016-0015 July 2016Salem

With Salem Unit 1 in a defueled condition during a planned refueling outage, anomalies were identified on baffle to former bolts while conducting a scheduled visual inspection of Reactor Vessel Internals. Due to the visual anomalies, PSEG commenced ultrasonic inspection of the baffle to former bolts to determine the extent of condition and determine a repair plan.

Based on initial results of the analysis of the ultrasonic inspection data received on May 03, 2016, this condition was determined to be reportable pursuant to 10 CFR 50.72(b)(3)(ii)(B), since the as-found conditions were not previously analyzed.

As a result of the inspection approximately 190 baffle to former bolts were identified as needing replacement.

05000311/LER-2016-00314 April 2016Salem

On 2/14/16 at 20:58 Salem Unit 2 automatically tripped from 100% power on Generator Protection.

The trip was initiated due to a Main Turbine trip caused by a Main Generator Protection signal. All emergency core cooling systems and emergency safeguards feature systems functioned as expected. The motor driven and steam driven auxiliary feed pumps started as expected on steam generator low level. Operators stabilized the plant in Mode 3 with decay heat removal via the main steam dump valves and auxiliary feed water system. Condenser vacuum remained available for the duration of the event. Operators also ensured a normal offsite electrical power lineup. Investigation identified a Stator Water Cooling valve leak dripping onto a relay, shorting the relay wiring terminations. This caused the turbine generator trip.

This report is being made in accordance with 10CFR50.73 (a)(2)(iv)(A), "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B)," for this event actuation of the Reactor Protection System and the Auxiliary Feedwater System.

05000311/LER-2016-0024 April 2016Salem

On 2/4/16 at 11:21, Salem Unit 2 automatically tripped from approximately 74% power. Power had been reduced at the beginning of dayshift to support a 500 KV transmission line outage. The reactor trip was due to a Main Turbine trip caused by a Main Generator Protection signal initiated by a main generator automatic voltage regulator (AVR) volts/hertz over excitation protection relay. All emergency core cooling systems and emergency safeguards feature systems functioned as expected. As found calibration data for the generator protection logic relay were found out of specification low. An evaluation determined the cause of the generator protection relay trip was poor manufacturing quality and/or shipping damage to an adjusting rheostat.

This report is being made in accordance with 10 CFR 50.73 (aX2Xiv)(A), "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B)," specifically automatic actuation of the Reactor Protection System and the Auxiliary Feedwater System for this event.

05000311/LER-2016-00121 March 2016Salem

On January 19, 2016, while reviewing outage data, plant staff recognized that anomalous data collected in October 2015, for the 21 Auxiliary Feed Pump time response loop resulted in failure to meet a surveillance, rendering that channel of Auxiliary Feedwater automatic actuation inoperable. In November 2015, the isolation valve for the pressure override defeat pressure transmitter was found closed. The pressure transmitter provides an input into the 21 AFW Pump run-out protection circuit. With the isolation valve closed, it would take longer to sense pump discharge head and consequently the opening of the auxiliary feed pump flow control valves would be slower than normal. This condition resulted in 21 Auxiliary Feedwater Pump loop time response greater than Technical Specification (TS) acceptance criteria. The failed channel was not recognized and the TS action was not taken, resulting in a condition prohibited by TS. The investigation revealed that the condition most likely existed since April 20, 2015, when maintenance activities were performed on the auxiliary feedwater pump discharge pressure transmitter. The isolation valve was opened and the surveillance was performed satisfactorily.

This report is being made in accordance with 10CFR50.73 (a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications.

05000311/LER-2015-0039 March 2016Salem

At 2136 on November 23, 2015, the Boron Injection Tank (BIT) relief valve 2SJ10 exhibited increased seat leakage during the performance of troubleshooting to determine the cause of low BIT pressure. The increased seat leakage from 2SJ10 initiated a Reactor Coolant System (RCS) leak greater than 10 gallons per minute (gpm). Technical Specification (TS) 3.4.7.2.b action b was entered for RCS unidentified leakage greater than 1 gpm.

The BIT was isolated at 2137 and the leakage was stopped. Isolation of the BIT resulted in loss of the high head safety injection flow path for both trains of high head safety injection, requiring entry into TS 3.0.3.

This event was caused by ineffective use of internal operating experience in the decision making process to reuse the 2CV141, which had been installed on the discharge of the 23 positive displacement charging pump, as a suitable replacement for the 2SJ10 during 2R21.

This report is being made in accordance with 10CFR50.73 (a)(2)(v)(D) "Any event or condition that could have prevented the fulfillment of the safety functions of structures or systems that are needed to mitigate the consequences of an accident"

05000272/LER-2015-00729 January 2016Salem

On October 28, 2015, Salem Unit 1 was in Mode 1 operating at 100 percent power and Unit 2 was in Mode 6 with fuel movement in progress. The Salem Unit 1 Control Room Emergency Air Conditioning System (CREACS) train was in single train filtration mode supplying air to the common Control Room due to outage activities on Unit 2.

At 0628, the Salem Unit 1 CREACS train was declared inoperable due to failure of its charcoal filter surveillance test, resulting in both units CREACS trains being inoperable. Unit 1 entered Technical Specification (TS) 3.0.3, and Unit 2 suspended fuel movement to comply with TS 3.7.6, Action c.

At 0950, the Unit 2 CREACS train was returned to service and aligned to single train filtration mode supplying air to the common Control Room. Unit 1 exited TS 3.0.3, meeting the requirement of TS 3.7.6.1, Action a., to restore a single train of CREACS to operable.

This event was caused by less than adequate procedure guidance and ownership of the surveillance activity by the maintenance shop responsible for performing the work. The direct cause of the TS entry was the filter failure due to aging.

This event is reportable under 10 CFR 50.73(a)(2)(i)(B), as a condition prohibited by the plant's TS, and 10 CFR 50.73(a)(2Xv)(D), as a condition that could have prevented the fulfillment of a safety function.

05000272/LER-2015-00623 October 2015Salem

On August 26, 2015, a fire scenario was identified that could cause spurious operation of the Pressurizer Power Operated Relief Valves (PORVs) and also prevent the ability to manually close the associated PORV Block Valves. The fire scenario would invalidate the assumptions in the Safe Shutdown Analysis with respect to Reactor Coolant System Inventory and Pressure Control.

The cause of this issue was a failure to validate previous corrective actions to ensure compliance with 10 CFR 50, Appendix R. Corrective actions include the development and installation of plant modifications to restore compliance with 10 CFR 50, Appendix R.

This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii)(B), "Any event or condition that results in the nuclear power plant being in an unanalyzed condition.

05000272/LER-2015-00119 October 2015Salem

(Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) On 1/16/15, irradiated fuel was being moved in the Salem Unit 1 Fuel Handling Building (FHB) in support of fuel sipping activities. Fuel Handling Building Ventilation system operability was being monitored via a camera focused on the local indication which was displayed in the Control Room as permitted by station procedures.

While fuel was being moved, the local differential pressure indicator oscillated between negative and slightly positive. Fuel movement was not secured as required by Technical Specification 3.9.12 when negative building differential pressure (D/P) was not maintained A causal evaluation determined that FHB equipment issues led to the FHB D/P alarm being locked in. A separate causal evaluation determined that Operations control room crew members failed to apply fundamentals of control board awareness and effective crew teamwork. Operating crew expectations were reinforced with respect to the assignment of personnel for continuous monitoring of equipment, and the underlying equipment deficiencies were corrected.

This event is reportable under 10 CFR 50.73 (a)(2)(i)(B) as an operation or condition which was prohibited by the plant's Technical Specifications.

05000311/LER-2015-0022 October 2015Salem

On 8/05/15, at 1539, Salem Unit 2 experienced an automatic reactor trip. The cause of the reactor trip was due to a trip of the 21 Reactor Coolant Pump (RCP) causing a 21 Reactor Coolant Loop low flow condition.

The 21 RCP breaker tripped as designed when the 2B Auxiliary Power Transformer (APT) infeed breaker to the 2H 4 kilovolt (kV) Non-Vital Bus opened. The root cause evaluation did not identify a definitive cause.

However the most probable cause of the 2H 4 kV Non-Vital Bus trip was due to a ground fault on the 21 Heater Drain Pump (HDP) motor that was not isolated by its associated neutral overcurrent relay. An automatic start of the Auxiliary Feedwater (AFW) system occurred as expected following the reactor trip due to low steam generator water levels.

Corrective actions include replacement of the 21 HDP motor and its neutral overcurrent relay.

This event is reportable under 10 CFR 50.73 (a)(2)(iv)(A) as an event that resulted in an automatic actuation of the reactor protection system and actuation of the AFW system.

05000272/LER-2015-00519 August 2015Salem

On June 22, 2015, at 1406, control room operators were notified by chemistry personnel that the Salem Unit 1 Containment Spray Additive Tank Sodium Hydroxide (NaOH) concentration was less than the minimum concentration by weight as required by Technical Specification (TS). Salem entered TS Action Statement 3.6.2.2.a for low NaOH concentration in the Containment Spray Additive Tank. On June 23, 2015 at 1908, NaOH concentration was returned to its minimum required TS value and the plant exited TS 3.6.2.2.a.

This report is made in accordance with 10 CFR 50. 73(a)(2)(i)(B) for "Any operation or condition which was prohibited by the plant's Technical Specifications ... " and 10 CFR 50.73(a)(2)(v) for "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: ... (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.

05000272/LER-2015-00419 June 2015Salem

Salem Unit 1 exceeded its Technical Specification (TS) allowed outage time for one channel of Overtemperature Delta-T (OT Delta-T) due to inadequate post-maintenance testing following replacement of a Power Range Nuclear Instrument (PR NI) upper detector cun-ent meter.

Corrective actions include an extent of condition review of other PR NI system post-maintenance testing and review of the post- maintenance testing procedure to identify deficiencies.

This report is made in accordance with 10 CFR 50.73 (a)(2)(iXB), "Any operation or condition which was prohibited by the plant's Technical Specifications...

05000272/LER-2014-0068 April 2015Salem

On October 19, 2014, at 2051, while performing a unit shutdown in preparation for its twenty-third refueling outage, Salem unit 1 control room operators initiated a manual reactor trip at approximately 20 percent reactor power. The manual reactor trip was inserted due to concerns with the 1 B Main Power Transformer, which had been in service with identified oil leakage. All control rods fully inserted on the trip.

The auxiliary feedwater system actuated as designed in response to low steam generator levels. Decay heat removal was via the steam dumps to the main condenser. The plant was stabilized in Hot Standby.

Operators failed to recognize parameters on 1 B Main Power Transformer which would have required them to enter an Adverse Condition Monitoring Plan. Entry into the plan would have required them to perform a fast load reduction and remove the main turbine from service at 40 percent power. Salem Generating Station will fully integrate procedure use and adherence principles and behaviors throughout Operations.

This report is made in accordance with 10 CFR 50.73 (a)(2)(iv)(A), "Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B) ... "for a manual reactor trip and for automatic actuation of the auxiliary feedwater system.

05000311/LER-2015-00119 March 2015Salem

On 1/20/15, a step change occurred on the high differential band indicator of the Axial Flux Difference recorder. Subsequent troubleshooting of the recorder on 1/26/15 identified a failed isolator that provides input to Channel 4 of Overtemperature Delta -T logic circuitry. This failure rendered Channel 4 of the Overtemperature Delta-T protection circuit inoperable for a period of approximately five and a half days without being placed in a tripped condition. The isolator was replaced and the Overtemperature Delta-T Channel was returned to service.

This report is made in accordance with 10 CFR 50.73 (a)(2)(i)(B), "Any operation or condition which was prohibited by the plant's Technical Specifications...

05000272/LER-2014-00521 October 2014Salem

On August 27, 2014, the 11 Safety Injection (SI) pump was being tagged out for planned maintenance. At 0243, the 11 SI pump was declared inoperable when its 4 kilovolt (kV) supply breaker was racked out. At 0248, the 12 SI pump failed to start on demand when operators were attempting to fill the 14 SI Accumulator. The unit entered Technical Specification (TS) 3.0.3 for inoperability of two SI pumps. At 0301, the 11 SI pump was realigned, tested and returned to service and the unit exited TS 3.0.3. The apparent cause of the 12 SI pump failure to start was due to a failure of the spring release solenoid in the pump's breaker.

This report is made in accordance with 10 CFR 50.73 (a)(2)(v)(D), "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed.

to: ...(D) Mitigate the consequences of an accident" for inoperability of both SI Pumps.

05000272/LER-2014-0034 June 2014Salem

On April 13, 2014 at 2113, Salem Unit 1 experienced an automatic reactor trip. The direct cause of the reactor trip was a generator lockout resulting from a main generator transformer overall differential relay trip. All control rods fully inserted on the trip. All three Auxiliary Feedwater (AFW) pumps started as expected in response to low steam generator levels and decay heat was removed by the steam dumps to the main condenser. Operators entered the emergency procedures for the plant trip and stabilized the plant in Mode 3 (HOT STANDBY).

The Main Generator Overall Differential Relay tripped due to a failed wiring termination on the C Phase Neutral Generator Current Transformer. The failed wiring termination was repaired.

This report is made in accordance with 10 CFR 50.73 (a)(2)(iv)(A), "...any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2Xiv)(B)...

due to an automatic reactor trip and actuation of the AFW system.

05000272/LER-2014-0024 June 2014Salem

On April 8, 2014 at 2112, Salem Unit 1 control room operators received an alarm indicating failure of the 12 Essential Controls Inverter. At 2113, operators observed indications of a trip of the 11 Steam Generator Feedwater Pump (SGFP). Operators manually initiated a Main Turbine automatic runback by manually tripping the 11 SGFP in accordance with procedures. At 2114, operators manually tripped the reactor due to a low level in the 13 Steam Generator (SG). All control rods fully inserted on the trip. All three Auxiliary Feedwater (AFW) pumps started as expected in response to low SG levels and decay heat was removed by the steam dumps to the main condenser. Operators entered the emergency procedures for the plant trip and stabilized the plant in Mode 3 (HOT STANDBY).

The cause of this event was due to a loss of power to the 11 SGFP speed probes. A ground on the electrical bus providing power to the speed probes was repaired.

This report is made in accordance with 10 CFR 50.73(a)(2)(iv)(A), "any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B)" for a manual reactor trip and automatic actuation of the AFW system.

05000272/LER-2014-00110 March 2014Salem

On January 7, 2014, Salem determined that control of the Turbine Driven Auxiliary Feedwater (TDAFW) pump enclosure High Energy Line Break (HELB) barrier door had been inadequate for work which was performed in the room on October 3, 2013. The cause of this event is attributed to an organizational failure to ensure that guidance provided in plant HELB program procedures contained sufficient justification for compensatory actions used for barrier impairments.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(ii)(B) as "The nuclear power plant being in an unanalyzed condition that significantly degraded plant safety"; 10 CFR 50.73(a)(2)(v) as "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) Remove residual heat, and (D) Mitigate the consequences of an accident"; and 10 CFR 50.73(a)(2)(vii) as "Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to: (B) Remove residual heat, and (D) Mitigate the consequences of an accident.

05000272/LER-2004-0033 September 2004Salem

On June 2, 2004 at approximately 1230 hours, the 1SVV26 valve (service water to the turbine building isolation valve) was declared inoperable as a result of investigation into an abnormal condition with Service Water in the turbine building. A plant shutdown was initiated in accordance with Technical Specification 3.6.1.1 "Primary CONTAINMENT INTEGRITY".� On June 2, 2004, non-licensed operations personnel identified an abnormal condition in the control of the turbine building service water pressure. During a routine tour of the turbine building while returning Unit 1 from its sixteenth refueling outage, the operator noticed that the temperature in the number 11 main turbine lube oil heat exchanger was approximately 110 degrees F.� Further investigation revealed that the service water to the turbine building regulating valve (1ST1) was full open with only 72 psig in the service water turbine header downstream of 1ST1. These conditions, low pressure and high temperatures, were not normal for the plant conditions at the time.

Further troubleshooting indicated that the service water to the turbine building isolation valve (1SW26) had been improperly installed and its motor operator improperly set up. The root cause of the event is attributed to failure to follow the process of match marking to assure that the actuator was properly installed with respect to the valve position.� Corrective actions taken were: the valve actuator was removed, the valve disc was rotated 180 degrees, and re-tested satisfactorily. Additional corrective actions will include revising the applicable installation procedures, including lessons learned for similar activities in upcoming outages and evaluating the event for inclusion in the continuing training program. This event is being reported under the requirement of 10CFR50.73(a)(2)(i)(A), as a completion of a plant shutdown to comply with Technical Specifications.

05000311/LER-2004-00311 June 2004Salem

On April 12, 2004, at approximately 10:33, the Control Room Emergency Air Conditioning System was placed in a condition where it did not comply with its design basis for post LOCA mitigation. During maintenance of the Salem Unit 1 Solid State Protection System, a safety injection signal was generated.

As a result of the invalid safety injection signal on Unit 1, the Control Room Emergency Air Conditioning System actuated to its accident pressurized mode alignment, in which the Salem Unit 1 emergency intake air dampers were isolated and the Salem Unit 2 dampers opened. In this configuration, Salem Unit 2 was in a condition where it did not comply with its design basis for post LOCA mitigation. The Salem dose analysis performed to meet the requirements of the General Design Criterion (GDC) 19 states that with only one train of the Control Room Emergency Air Conditioning System available at the start of a design basis LOCA, the make up air supply to pressurize the control room envelope must be supplied by the non­ accident Unit's emergency outside air intake. The apparent cause of this event was a defective universal logic card in the Solid State Protection System (SSPS). When this card was moved from one position to another in the SSPS cabinet, a safety injection signal from Unit 1 Train 'B' occurred. Corrective actions taken were: (1) The defective card was replaced, and (2) A full functional test procedure on Train B was performed satisfactory. This condition is reportable under 10 CFR 50.73(a)(2)(v)(D).

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (1.2001r� 1 DOCKET (2)FACILITY NAME (1) LER NUMBER 6) PAGE (3)