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At 0938 (EDT), the watertight door in the Unit 1 Inboard Service Water Penetration Area was unable to be closed. The watertight door serves as a High Energy Line Break (HELB) and Medium Energy Line Break (MELB) barrier between the mechanical penetration room and the service water penetration room. A HELB/MELB event occurring in a room with its barrier door open could adversely affect equipment in an adjacent room. Consequently, a HELB/MELB event could have rendered equipment in the adjacent room inoperable. At 1005, station maintenance was able to successfully close and latch the door restoring the barrier. This event is being reported under the requirements of 10CFR50.72(b)(3)(ii)(B) as, 'the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' An ENS notification is required if an unanalyzed condition occurred within 3 years of the date of discovery even if the event is not on-going at the time of discovery. The licensee notified Lower Alloways Creek Township and the NRC Resident Inspector.

• * * RETRACTION FROM JOHN OSBORNE TO JOHN SHOEMAKER AT 1545 EST ON 12/12/16 * * *

The purpose of this notification is to retract event report number 52294 made on 10/11/2016 at 1648 (ET). Previously PSEG reported that Salem Unit 1 was determined to be in an unanalyzed condition due to being unable to close the Unit 1 Inboard Service Water Penetration Area Water Tight Door. The watertight door was reported to serve as a High Energy Line Break (HELB) and Medium Energy Line Break (MELB) barrier between the mechanical penetration room and the service water penetration room. Subsequent review identified that the condition did not meet the reporting criterion. Engineering evaluation determined that the service water penetration room had been previously evaluated for the impact due to a HELB event occurring in the adjacent mechanical penetration area with the watertight door open and that the event would not impact the operability of the service system. Engineering evaluation also determined that the Salem Unit 1 Design Basis does not require analysis of a MELB event occurring in the service water penetration room. This is due to the timing of the Unit 1 operating license issue date. At the time of the issuance of the Unit 1 operating license analysis of a MELB event was not required. Therefore an unanalyzed condition for a MELB event did not exist. Additionally while the door was difficult to close, it was able to be closed and dogged in a reasonable time interval (27 minutes), therefore, any potential internal flooding which would have been detected immediately by the attendant with required actions taken to close the door and isolate the leak rapidly. Therefore PSEG is retracting the notification made under 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector was notified of this retraction by the licensee. Notified R1DO (Schroeder).

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 an unanalyzed condition. Preliminary investigation has revealed that the condition most likely existed since April 20, 2015, when maintenance activities were performed on the auxiliary feedwater pump discharge pressure transmitter. Consequently, there were multiple instances when one of the other auxiliary feedwater pumps was removed from service, thus creating a condition which did not meet the accident analysis assumptions for auxiliary feedwater flow initial response. This event is being reported under the requirements of 10 CFR 50.72(b)(3)(ii)(B) as 'the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' This condition was corrected on November 20, 2015. The auxiliary feedwater pump discharge pressure transmitter instrument isolation valve was inadvertently left closed after the April 20, 2015 maintenance. The licensee has notified the NRC Resident Inspector and will notify the State of New Jersey, State of Delaware, and the local township.

• * * RETRACTED ON 3/10/16 AT 1647 EST FROM JACK OSBORNE TO DONG PARK * * *

This event is being retracted. An engineering review determined that while the auxiliary feedwater flow loop response time test results did not meet the procedural acceptance criteria, the accident analysis assumptions remained valid. The ATWS (Anticipated Transient Without Scram) was the limiting accident, and the loop response time results were still bounded by the existing analyses. The failed loop response time did result in a condition prohibited by TS which will be reported in a Licensee Event Report (LER). This condition has been documented in the licensee's Corrective Action Program. The licensee has notified the NRC Resident Inspector. Notified R1DO (Dimitriadis).

At 0226 (EDT on 10/12/14), the 25 Service Water Pump Traveling Screen differential pressure transmitter high side valve, 2LD2729-HIV was discovered closed while performing the monthly bubbler blow down activity. The associated 25 Service Water Pump was operable at this time. The differential pressure transmitter high side valve, 2LD2729-HIV, in the closed/discovered position would have prevented the operation of the 25 Service Water Traveling Screen due to high differential pressure. The 25 Service Water Traveling Screen needs to be operable to support 25 Service Water Pump operability. 25 Service Water Pump Traveling Water Screen was restored to operable after differential pressure transmitter high side valve, 2LD2729-HIV was reopened. The station subsequently verified all Unit 1 and 2 high side and low side differential pressure transmitter valves positions were correct. At 0600 (EDT on 10/12/14), it was identified that the last manipulation of differential pressure transmitter high side valve, 2LD2729-HIV was on 9/7/14. Based on the last known manipulation it is assumed that differential pressure transmitter high side valve, 2LD2729-HIV remained closed from that time until the condition was discovered. Review of other activities performed from 9/7/2014 to present determined that surveillance testing of 21 Service Water Pump resulted in 21, 22, and 23 Service Water Pumps being inoperable on 9/18/2014 for several hours. During that surveillance, combined with the mis-positioned instrument valve on 25 Service Water Pump, five of the six Service Water Pumps would have been inoperable which may have prevented the fulfillment of a safety function. This event is being reported under the requirements of 10CFR50.72(b)(3)(v)(B) as 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of systems that are needed to remove residual heat.' The licensee has notified the NRC Resident Inspector. No one was injured as a result of the failure of 25 Service Water Traveling Screen inoperability.

• * * RETRACTION FROM ROBERT CORDREY TO DONALD NORWOOD AT 1543 EST ON 12/2/2014 * * *

The purpose of this call is to retract event number 50527. On October 12, 2014, at 1307 EDT, notification was made to the NRC Operations Center by Salem Unit 2 reporting a condition under the requirements of 10CFR50.72(b)(3)(v) as 'Any event or condition that at the time of discovery could have prevented the fulfillment of a safety function of structures or systems that are needed to: ...(D) Mitigate the consequences of an accident.' Subsequent to this report, Salem Unit 2 determined that the condition in which the 25 Service Water (SW) pump may have been considered inoperable, in conjunction with the other SW loop being out of service, was not occurring at the time of discovery and was thus not reportable under the reporting requirement of 10CFR50.72(b)(3)(v). Further analysis of the event determined that the valve mis-positioning of the 25 SW pump traveling screen differential pressure transmitter which defeated the auto-start function of the traveling screen, did not impair the safety function of the traveling screen to minimize carryover of debris to the suction of the 25 SW pump. Thus, during the time in question, the 25 SW pump remained operable which maintained one loop of SW operable during the period of time the other service water loop was removed from service for pump testing. One loop of SW with two pumps operable from different safety related buses is capable of performing the safety function of the SW system during a design basis accident. This event is not considered a safety system functional failure and is not reportable to the NRC under the requirements of 10CFR50.72(b)(3)(v) or 10CFR50.73(a)(2)(v). The NRC Resident Inspector has been notified. Notified R1DO (Dentel)

Salem Generating Station made a 15 minute notification of a chemical discharge to the State of New Jersey Department of Environmental Protection at 1806 (EST). The Salem Non-Rad Waste Chemical Treatment Building sump overflowed out of the building to a catch basin that discharges to the Delaware River. Approximately 100 gallons of chlorinated water was reported to the State of New Jersey as being discharged, which was terminated at 1806 (EST). There were no personnel injuries associated with this event. There was no impact to any Salem Station Safety-Related systems and all Safety-Related systems are available. Investigation into the cause of the event is in progress. The licensee notified the NRC Resident Inspector and Lower Alloways Township.

• * * RETRACTION FROM DAN MCHUGH TO JOE O'HARA AT 1031 ON 1/17/12 * * *

On 1/14/2012 Salem Generating Station made a 15 minute notification of a chemical discharge to the State of New Jersey Department of Environmental Protection at 1806. The Salem Non-Rad Waste Chemical Treatment Building sump overflowed out of the building to a catch basin that discharges to the Delaware River through DSN 488. Subsequent investigation has indicated that the spill was entirely contained within the onsite storm drainage system and that there was no discharge to the Delaware River. The storm drain system was plugged, flushed and pumped out for appropriate disposal of the waste water. Additionally, the spill to the ground was cleaned up within 24 hours. The licensee will notify the NRC Resident Inspector and has notified the New Jersey Department of Environmental Protection.

Salem Unit 1 was placed in a configuration that affected the ability to mitigate the consequences of an accident due to an inadvertent actuation of the common control room emergency air conditioning system (CREACS). CREACS was actuated as a result of an invalid Control Room air intake duct radiation monitor initiated on April 4, 2008 at 1259 hours. Salem Unit-2 is currently defueled. Salem Unit 1 is in Mode 1 at 100% power. Unit 1 has two shutdown LCOs in effect. The first is for the CREACS system, which is shared between Unit 1 & 2, being aligned for single train operation with the Unit 2 CREACS train out of service per LCO 3.7.6. The second shutdown LCO is for two outside air intake dampers being inoperable for scheduled maintenance. With Unit 2 having an invalid Radiation Monitor signal, the CREACS system actuates in accident pressurized mode. This mode of actuation starts the CREACS fans, isolates the Control Room Envelope from the normal control room ventilation system and aligns the two sets of CREACS outside air intake dampers. With a Unit 2 Radiation Monitor signal, the Unit 2 CREACS intake dampers close and the Unit 1 CREACS intake dampers open. These damper positions are locked in until manually reset. With only one train of CREACS operable, the dose analysis indicates that the requirements of General Design Criterion (GDC) 19 can only be met during the worst case design basis accident if the Unit 1 CREACS intake dampers are closed and the Unit 2 CREACS intake dampers open. Therefore, until the CREACS intake dampers were reset and realigned, Salem Unit 1 would not have been able to mitigate the consequences of an accident and is reportable in accordance with 10CFR50.72(b)(3)(v)(D). The CREACS system actuation was reset after the failed radiation monitor (1R1B ch II) was removed from service and the dampers were re-aligned to their pre actuation alignment at 1316 hours, restoring Salem Unit 1 to within the assumptions of the dose analysis. Total duration in the condition was 17 minutes. The only piece of major equipment out of service on Salem Unit 1 is the 15 Service water pump which is out of service for scheduled maintenance. The Licensee notified the NRC Resident Inspector.

• * * RETRACTION PROVIDED BY ERIC POWELL TO JASON KOZAL ON 04/17/08 AT 2113 * * *

On April 4, 2008 Salem Unit 1 was placed in a configuration that was contrary to the current dose analysis of record due to an inadvertent actuation of the common control room emergency air conditioning system (CREACS). CREACS was initiated as a result of an invalid actuation of a Control Room air intake duct radiation monitoring channel (1R1B ch II). At the time of the actuation, the Unit 2 train of CREACS was out of service due to scheduled maintenance leaving only the Unit 1 CREACS train operable. Unit 1 was at 100% power and Unit 2 was defueled. With one train of CREACS out of service at the start of an accident the dose analysis of record requires that the CREACS emergency intake dampers for the unit having the accident to close and for the opposite units emergency intake dampers to open. The actuation of the radiation monitoring channel 1R1B channel II caused the Unit 1 emergency intake dampers to open which was contrary to the dose analysis of record. Subsequent to this event, an evaluation was performed utilizing the assumptions of the dose analysis of record with the exception of the actual measured engineered safety feature system leakage outside containment and the atmospheric dispersion factors (x/Q) associated with the Unit 1 CREACS intake. Based upon this evaluation, the CREACS system with one filtration train operable and the emergency intakes open on Unit 1 would have been able to maintain doses to the Control Room operators below the limits of GDC-19 and the dose analysis of record. Since the CREACS system was capable of performing its accident mitigation function, this event is being retracted. The licensee will notify the NRC Resident Inspector. Notified R1DO (Schmidt)

This is an 8-hour notification being made to report exceeding the design basis for reactor coolant leakage outside of containment. The normal daily RCS leakrate was completed at 2345 on 2/15/06. This leakrate indicated a step change in unidentified leakrate to .8 gpm from .09 gpm. This leakage value is within the 1 gpm allowed by Technical Specifications. Investigation is on going, and the source of the leak has not been determined at this time, however preliminary conclusion is that the leakage is outside of containment and related to the centrifugal charging pumps. The design requirement ECCS leakage outside of containment is 3840 cc/hour ( .1 gpm) to support GDC-19 limits for control room habilitability. No safety system actuation occurred or were required. No injuries have occurred due to this event. The licensee has performed troubleshooting to identify the source of the leak and to narrow down the portion of the charging system where they believe the leak to be located. The licensee believes the leak is in a relief valve or hard pipe system to the waste tanks, and that the leak is not external to the system. The licensee notified the NRC Resident Inspector.

• * * RETRACTION FROM PAUL ABBOTT TO JOE O'HARA AT 2119 ON 4/6/06 * * *

On February 16, 2006, PSEG made an 8-hour report (event number 42343) in accordance with 10CFR50.72(b)(3)(v) because of exceeding the design basis limits for ECCS leakage outside containment. The leakage was determined to be approximately 0.8 gpm with a limit of 3840 cc/hour. Further investigation into this event determined that the leakage was not in the ECCS recirculation flow path. The cause of the elevated RCS unidentified leakrate was the failure of the automatic three-way high level divert valve (2CV35), which prevents a high level from occurring in the Volume Control Tank (VCT). This valve failed to fully isolate flow to the Chemical Volume Control (CVC) Hold-Up Tanks (HUTs) after VCT level dropped below the divert setpoint of 77% . This conclusion is based on the VCT level being at the divert setpoint a number of times during the shift, the increased use of the valve to control level, and valve performance to fully isolate flow to the CVC HUT'S. Because this flow path is automatically isolated on a Safety Injection signal by other means (valves), the leakage through this valve is not considered part of the ECCS recirculation flow path and therefore is not included in the calculation for ECCS leakage outside containment. The event is being withdrawn. The licensee will notify the NRC Resident Inspector. R1DO(Finney) notified.

This is an 8-hour notification being made to report exceeding the design basis for the reactor coolant leakage outside of containment. An unidentified reactor coolant system (RCS) leak of 1.0 gpm on Salem Unit 1 occurred on 7/28/04 at 2100 hours. It was discovered (that) a pressure instrument weld from 13 charging pump (PDP) discharge piping failed in the auxiliary building. Tech Specs for RCS unidentified leakage was entered. Station abnormal operating procedures (ABRC-001 RCS Leak) were entered and leakage monitoring program referenced. The leak was isolated closing the root valve of the instrument line (discharge pressure transmitter) by 2135 hours. The RCS leakage was outside of containment and was greater than design of 3840 cc/hr. This leakage would have exceeded the GDC-19 limits for control room habitability during the time of the leak initiation and leak isolation. 13 charging pump remained in service and the leak is locally isolated. The tech spec for RCS unidentified leakage was exited and ASME tech spec for the failed weld was entered. There were no unusual or unexpected responses. All systems functioned as required. There were no personnel injuries. A maintenance tech in the auxiliary building notified control room of a leak about the same time that control room personnel noticed that the Volume Control Tank level was decreasing approx. 0.3gpm. The area around the Positive Displacement Pump (PDP) was roped off and the leak was isolated. Airborne levels in the roped off area was 8000 cpm (from 2130 to 2200 hours). Nobody was evacuated from auxiliary building. At 0040 hours airborne samples were within normal limits. The area around # 13 PDP is still roped off until the surface contamination is cleaned up. Lower Alloways Creek Township will be notified of this by the licensee. The NRC Resident Inspector will be notified of this event by the licensee.

• * * RETRACTION FROM SAUER TO CROUCH AT 1543 HRS. EDT ON 9/21/04 * * *

The following retraction was obtained from the licensee via e-mail: On July 28, 2004, it was discovered that a pressure instrument weld from 13 charging pump developed a leak into the auxiliary building. An 8-hour call to the NRC was made for exceeding the ECCS leakage outside containment (Event Number 40904). The leak was self-revealing, in that the area radiation detectors as well as the fire protection sensors detected it. The leak was identified and isolated by closing the manual root valve for the instrument line in approximately 35 minutes from discovery. Section 5 of Appendix A of Regulatory Guide 1.183 states that engineered safeguards feature (ESF) systems that recirculate sump water outside of the primary containment are assumed to leak during their intended operation. The guidance indicates that this release source includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar components. The guidance further indicates that this release source may also include leakage through valves isolating interfacing systems. The leak that was identified and isolated shortly after discovery was a piping system failure and not the type of leakage that is considered to be a release source for which the radiological consequences should be analyzed. Therefore, further review of the event indicated that identified leak did not challenge the limits of 10CFR100 for offsite releases and 10CFR50 Appendix A General Design Criterion 19 (GDC-19), thus event report 40904 is withdrawn. The licensee has notified the NRC Resident Inspector of the retraction. The Headquarters Operations Officer notified R1DO (Caruso).

This is an 8-hour notification being made to report that a valid ESF Auxiliary Feed actuation occurred. Salem Unit 1 is in mode 5 with RHR providing shutdown cooling. '11' Aux Feed Pump was in service to fill '13' and '14' steam generators for wet lay-up conditions. Actual levels were low in both '13' and '14' steam generators but jumpers were installed on steam generator narrow range level channel III and IV to prevent an ESF actuation. On 4/03/04 at 1050 AM, a breaker effecting reactor protection system channel III was cleared and tagged as part of preparations to remove 1C 4KV vital bus from service. The actual low level in '13' and '14' steam generators along with the power loss to channel III caused 2 out of 3 logic to be satisfied and initiated an AFW actuation. '11' Auxiliary Feed Pump remained running. '12' Auxiliary Feed Pump auto started. '13' Auxiliary Feedwater Pump was removed from service prior to the actuation and did not start. The '13' and '14' steam generators continued to fill. '11' and '12" Auxiliary Feed Valves (11AF21 and 12AF21) were closed and no level rise was observed. The breaker was restored and reactor protection system channel III was placed back in service. All auto start signals cleared after power was restored and the '12' Aux Feed Pump was stopped at 1101. There were no unusual or unexpected plant response from the actuation. All safety systems and equipment performed as expected. Entry into mode 6 is expected this afternoon and progress of the refueling outage is expected to continue. There were no personnel injured. The licensee will inform the Lower Alloways Creek Township (LAC) and has informed the NRC Resident Inspector.

• * * RETRACTION ON 05/05/04 AT 1252 EDT FROM S. SAUER TO A. COSTA * * *

On April 3, 2004 at 1410 PSEG made an 8 hour notification to report a valid ESF actuation of the auxiliary feedwater system (Event Number 40639). At the time of the report 11 Auxiliary feedwater pump was in service to fill the 13 and 14 steam generators for wet lay-up. The levels on those generators were low (as previous plant condition had demanded) and the jumpers had been installed in the level detectors to prevent the automatic start of the auxiliary feedwater pumps. 12 auxiliary feedwater pump was out of service. The steam driven auxiliary feedwater pump was tagged out of service. Core heat removal was being provided by the Residual Heat Removal System. On April 3, with the steam generator level being carried below the low level setpoint, in accordance with procedures, as a result of other activities associated with the refueling outage the installed jumpers were removed causing the auto start of the 12 pump. Subsequent investigation Into this event and further review of NUREG 1022 has determined that the condition described above is not reportable under the requirements of 10CFR50.72(b)(3)(iv) or 50.73(a)(2)(iv). As stated In NUREG 1022 the intent of reporting under this paragraph is '..to report actuations of systems that mitigate the consequences of significant events .. The Staff does not consider this to include single component actuation because single components of complex systems, by themselves usually do not mitigate the consequences of significant events.' Furthermore valid signals are defined as ' those signals that are Initiated in response to actual plant conditions .. Satisfying the requirements for initiation of a safety function of the system.' (emphasis added on safety function). In this particular event the required Safety Function to maintain the core cooled and decay heat removal was being accomplished by the Residual Heat Removal System and it remained unaffected throughout this event. The plant was in a condition where the steam generators in conjunction with the Auxiliary Feedwater System were not part of the ultimate heat sink or a principal means to remove decay heat. The Auxiliary Feedwater System was only functional and available to provide a means to place the steam generators in wet lay-up in support of outage activities. The 11 pump was already in service providing for this non-safety related function. Thus the auto start of the 12 pump was not as a result of a valid signal for a significant event that required initiation of a mitigating function; e.g. an ESF actuation. Therefore this event was not reportable under 10CFR50.72 or 50.73, as per the guidance provided in NUREG 1022. The licensee notified the NRC Resident Inspector. Notified R1DO(Della Greca).