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At 09:30 EDT, Unit 1 was performing the HPCI Quarterly Flow verification surveillance. Shortly after the initiation of the system an abnormally loud bang was heard. System flow of approximately 5200 gpm and discharge pressure of approximately 1300 psi was achieved at approximately 09:33. Approximately 4 seconds after reaching rated system flow HPCI discharge pressure increased to approximately 1675 psi and system flow dropped to approximately 2700 gpm. HPCI had been declared inoperable at 08:35 EDT to perform the surveillance and will remain inoperable until the cause of the loss of system flow is corrected. Because HPCI is a single train ECCS (Emergency Core Cooling System) safety system, this event results in the loss of an entire safety function which requires an 8 hour ENS notification in accordance with 10CFR50.72(b)(3)(v) and the guidance provided under NUREG-1022, rev. 2. There are no other ECCS systems presently out of service. Reactor Core Isolation Cooling (RCIC) is fully operable and HPCI entered Tech Spec 3.5.1 (14 day Limiting Condition of Operation). All other ECCS systems are fully operable. The NRC Resident Inspector was notified of this event by the licensee.

• * * RETRACTION FROM WALSH TO CROUCH ON 10/02/03 @ 1548 EDT* * *

On 09/05/2003, PPL Susquehanna LLC made an ENS notification per 10CFR50.72(b)(3)(v) in response to an apparent loss of the HPCI (High Pressure Coolant Injection) safety function. In the event, results of HPCI Quarterly Flow Surveillance testing did not meet acceptance criteria established for the system. Investigation into the cause of the failed HPCI surveillance revealed a 360-degree weld crack on the HPCI Test Line to Condensate Storage Tank (CST) Valve, HV155F008. HV155F008 is not in the reactor vessel injection flowpath. The impact of this crack, which was located between the valve seat cage assembly and the valve body, was that the HPCI Test Line to CST valve was not capable of throttling over the full range of HPCI system flows. Valve performance became erratic at higher flows because the valve seat cage was lifted out of the valve body into the flow path, increasing system resistance, and preventing attainment of design flow in the HPCI test loop. While the ability to effectively test the HPCI system using the test return path to the CST was compromised, the HPCI injection flowpath to the reactor vessel was not adversely affected by the damaged valve. Accordingly, the HPCI system maintained full capability for providing sufficient coolant to the reactor vessel in the event of a small break loss-of-coolant accident. Because the HPCI safety function was not compromised by the identified test path obstruction, this ENS notification is being retracted. The Licensee has notified the NRC Resident Inspector. Notified R1DO (Cobey).

Appendix J Local Leak Rate Testing (LLRT) of the RCIC Steam Supply failed the Combined Main Steam Isolation Valves leakage limit of 300 scfh. The Combined MSIV leakage limit includes MSIV'S, MSL Drains, HPCI Steam Supply and RCIC Steam Supply. The volume between RCIC Steam Supply valves could not be pressurized therefore, the Minimum Pathway leakage limit is considered to be exceeded. This is also considered a failure of the Primary Containment Air Leakage Minimum Pathway leakage limit of 1.0 La. The identified degraded condition is reportable as a condition of the nuclear power plant, including its principle safety barriers being seriously degraded per 10CFR50.72(b)(3)(ii) requiring an 8-hr ENS notification. The RCIC Steam Supply line was tested successfully approximately two years ago during the last refueling outage. The licensee will inform the NRC Resident Inspector.

• • • • • • RETRACTED ON 4/14/2004 AT 1610 EST FROM FRY TO HEISSERER*****

At the time the original 8-hour ENS notification was made, it was declared that the Appendix J Local Leak Rate Test of the RCIC Steam Supply had failed the combined Main Steam Isolation Valves Minimum Pathway leakage limit of 300 scfh. The combined MSIV leakage limit includes MSIVs, MSL Drains, HPCI Steam Supply and the RCIC Steam Supply. The volume between the inboard and outboard RCIC Steam Supply valves could not be pressurized, therefore the Minimum Pathway leakage limit was considered to be exceeded. Subsequent to this event, during performance of an additional Local Leak Rate Test of the RCIC Steam Supply Penetration, it was demonstrated that the as-found Appendix J and combined MSIV Minimum Pathway leakage and the Primary Containment Air Leakage Minimum Pathway of 1.0 La were not exceeded as previously reported. The original Local Leak Rate Test was performed between the combination of two parallel inboard valves and the outboard valve. The subsequent test determined that the outboard isolation valve did not exceed Minimum Pathway criteria. The majority of the leakage was from the smaller inboard RCIC warm-up line isolation valve of the three RCIC Steam Supply valves. The post-maintenance test of the penetration measured leakage that was within acceptable limits. As such, the condition of the nuclear power plant, including its principle barriers, was not significantly degraded. The NRC Resident Inspector has been notified. Notified R1DO (B. McDermott).

At 22:10 on March 6, 2004 the Unit 1 Safety Parameter Display System (SPDS) was deenergized during planned refuel outage activities. It has been determined that it will not be restored within the required eight hours. Currently, restoration is scheduled for March 7, 2004 at 17:00. All required instrumentation is available in the control room. Since the Unit 1 SPDS computer system will be unavailable for greater than 8 hours, this is considered a Loss of Emergency Assessment Capability and therefore reportable under 10CFR50,72(b)(3)(xiii). Unit 1 is currently in Mode 5 for the planned 13th Refuel and Inspection Outage. The licensee notified the NRC Resident Inspector

• • • • Update at 2219 EST on 03/07/04 by Gordon Robinson taken by MacKinnon ****

SPDS is still out of service and it will be restored to service by 0600 EST on 03/08/04. R1DO (M. Shanbaky) notified. NRC Resident has be notified of the update by the licensee.

• * * UPDATE ON 03/09/04 @ 1122 BY GRANT FERNSLER TO C. GOULD* * *

As of 1117 the SPDS was energized and restored to service. The NRC Resident Inspector was notified. Reg 1 RDO(Cobey) was notified.

At 1040 hrs during setup for Unit 1 Division 1 LOCA/LOOP Surveillance testing, the controlling procedure required making the Loss of Power instrumentation for the 1A and 1C ESS buses inoperable. The specification has a 1 hour time limit for restoration, or the associated Diesel Generators must be declared inoperable. Due to delays during the setup of equipment the time requirements were not met, and the 'E' (Substituting for 'A') and 'C' Diesel Generators were declared inoperable at 1140 hrs. The Loss of Power instruments were restored, and the Diesel Generators declared operable at 1206 hrs. Unit 1 is in Mode 5 requiring only 2 diesel generators operable, therefore not impacted by the Loss of Power instrumentation inoperability. Also during the surveillance, two pump start timers failed to meet the required acceptance criteria. The 'A' ESW Pump timer actuated at 47.86 seconds (criteria; 36 sec. to 44 sec) and the 'C' ESW Pump timer actuated at 50.38 seconds (criteria: 39.6 sec. to 48.4 sec). With failure of the timers, proper loading on the Diesel Generators is not assured, and they were declared inoperable until the associated pump control breakers were opened. The 'E' Diesel Generator was declared inoperable at 1448 hrs, when the data analysis identified that the 'A' ESW pump timer did not meet acceptance criteria. The 'E' Diesel Generator was returned to operable at 1453 hrs when the DC Knife switches for the 'A' ESW pump were open. The 'C' Diesel Generator was declared inoperable at 1524 hrs, when the data analysis identified that the 'C' ESW pump timer did not meet acceptance criteria. The 'C' Diesel Generator was restored to operable at 1706 hrs after supported systems were realigned to prevent further loss of safety function and the DC Knife switches for the 'C' ESW pump were open. The Susquehanna safety analysis requires three operable Diesel Generators to safely shutdown the plant. Therefore with only two operable Diesel Generators, the condition requires an 8 hr ENS notification in accordance with 10CFR50.72(b)(3)(v)and (vi). All times referenced above are EDT. The licensee informed the NRC Resident Inspector.

* * * RETRACTION FROM R. FRY TO M. RIPLEY 1419 ET ON 05/27/04 * * * *

The following is a retraction of ENS Notification #40642 AND #40646: On April 5, 2004, PPL Susquehanna, LLC reported three events to the NRC associated with Unit 1 Division 1 LOCA/LOOP surveillance testing. In each case, the safety function of two (out of 4) emergency diesel generators was brought into question during the testing. In two of these events, the E (substituting for A) and C diesel generators were administratively declared inoperable per Technical Specifications. This was necessary because Loss of Power instrumentation for the 1A and 1C buses, rendered inoperable in support of the test procedure, was not restored within 1 hour. In the third event, the E (for A) and C emergency diesel generators were, again, declared inoperable when two ESW pump start timers failed to meet acceptance criteria thus bringing the proper start sequencing of emergency loads into question. Because the Susquehanna Safety Analysis requires three diesel generators to safely shutdown an operating unit, the potential compromise of the diesel generator safety function was a concern for Unit 2 in each instance. (NOTE: Unit 1 was in Refueling Mode 5 and was, therefore, not impacted by these events.) Accordingly, Control Room personnel conservatively initiated ENS reporting under 10CFR50.72(b)(3)(v) in response to the apparent loss of safety function for Unit 2. Subsequent engineering analysis has concluded that, although the declaration of diesel generator inoperability was correct and in accordance with Technical Specifications, fulfillment of these administrative actions did not result in an actual loss of safety function during any of the above described events. Details of the analysis show that the diesel generators were available to supply emergency power to the ESS buses and/or that sufficient redundant equipment was available to fulfill the safety function needs of Unit 2. These analysis conclusions provide the basis for retraction of the ENS reports made on April 5, 2004." The licensee notified the NRC Resident Inspector. Notified R1DO (R. Conte)

At 09:20 am, on February 9th, the control room declared both trains of Control Structure Emergency Outside Air Supply System (CREOASS) inoperable. The control room was notified of an inoperable boundary door that was not closed and latched and therefore would not have allowed CREOASS to perform its safety function of maintaining the Control Structure at a positive pressure if required. The door was closed and locked (at 11:15) to restore the habitability boundary. The door has since been repaired (adjustment of latch mechanism). Initial investigation determined that the door had been unlocked at approximately 09:20 this morning by an elevator vendor to perform maintenance. This is being reported as an event or condition that could have prevented the fulfillment of the safety function required to mitigate the consequences of an accident. CREOASS was not called upon to function during the period that the door was not operable. The licensee notified the NRC Resident Inspector.

• * * RETRACTION FROM LICENSEE (ROBINSON) TO NRC (HUFFMAN) @ 1328 EDT ON 4/8/05 * * *

The following information was provided by the licensee (licensee text in quotes): On 2/9/2005, PPL Susquehanna LLC reported a loss of safety function for the Control Room Emergency Outside Air Supply System (CREOASS) based on one of the Control Structure Habitability boundary doors being found partially open. Positive air pressure generated by the Control Structure HVAC System prevented the door closure mechanism from maintaining the door closed. The door was closed to restore the habitability boundary. The event was reported as a condition that could have prevented the fulfillment of the safety function required to mitigate the consequences of an accident. The Safety Function of the CREOASS is to provide adequate radiation protection to permit occupancy of the Control Room under accident conditions for the duration of the accident. The CREOASS accomplishes this by: 1) filtering the outside make-up air supply and 2) maintaining a positive pressure in the control structure habitability boundary to limit unfiltered air in leakage into the boundary. A subsequent engineering evaluation concluded that with the door closure mechanism providing a resistive force, CREOASS would have been able to maintain a positive pressure within the boundary and perform its safety function as described above. In conclusion, the reported condition did not result in a loss of safety function under 10CFR50.72 (b)(3)(v)(D) and is therefore being retracted. The NRC Resident Inspector and R1DO (Noggle) have been notified.

The following information was provided by the licensee via facsimile (licensee text in quotes): At 1500, on March 6, 2005, the Control Room declared both required divisions for three functions (Primary Containment Pressure, Primary Containment Hydrogen and Oxygen Analyzer, and Drywell Atmosphere Temperature) of Post Accident Monitoring Instrumentation (a Safety System) inoperable. The control room was notified of 'Non Quality' (non-Q) parts installed in both required divisions of a Post Accident Monitoring Instrumentation Recorder. The appropriate LCO Conditions were entered for one or more functions with two required channels inoperable. This equipment has passed all surveillance requirements and has been functional since installation. Plans are being developed to replace the non-qualified parts. This is being reported as an event or condition that could have prevented fulfillment of a safety function required to mitigate the consequences of an accident in accordance with 10CFR50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector.

• * * UPDATE ON 03/30/05 @ 1128 BY JIM HUFFORD TO CHAUNCEY GOULD * * * RETRACTION

The following information was provided by the licensee via facsimile (licensee text in quotes): On March 6, 2005, PPL Susquehanna, LLC reported a perceived loss of safety function for three functions of Post Accident Monitoring Instrumentation on Unit 1. The notification was made pursuant to the reporting requirements of 10CFR50,72(b)(3)(v)(d). Engineering analysis has subsequently determined that the use of non-Q fuse holders in the Quality application did not adversely affect the safety-related functions which they supported. The fuse holders were removed, subjected to testing, and determined to be electrically and functionally equivalent 'to Quality fuse holders maintained at the station. This analysis conclusion provides the basis for retraction of the ENS report of March 6. The NRC Resident Inspector was notified. The Reg 1 RDO (Jim Trapp) was informed.

At 0352 the 2C 125VDC battery charger failed. Fuses internal to the charger were found blown. Technical Specifications require restoration of the charger within 2 hrs or enter Mode 3 within the next 12 hrs and Mode 4 within the next 36 hrs. Investigation has not been able to restore the charger, and therefore the shutdown of the unit has commenced. Loads are currently being carried by the batteries, and personnel are monitoring battery voltage. Initial investigation has not identified any generic issues or any indication of tampering. This event is a Technical Specification required shutdown, and is reportable as a four hour ENS Notification under 10 CFR 50.72(b)(2)(i). Additionally, this is a voluntary notification due to an anticipated press release associated with the downpower towards unit shutdown. There is evidence of degraded insulation on wires inside the affected battery charger. Battery loads are very small with an initial estimate of battery life at 80 hours. All safety systems are operable. There is no impact on Unit 1. Unit 2 is expected to reach Mode 3 at 1720 on 4/10/05. The NRC Resident Inspector has been notified. State (FEMA) will be notified.

• * * UPDATE PROVIDED BY LICENSEE (HUFFORD) TO NRC (HELD) AT 2030 ON 4/10/05 * * *

At 17:28 the plant completed a normal reactor shutdown and entered Mode 3 in accordance with plant procedures to fulfill the requirements of Technical Specification 3.8.4. There were no ECCS initiations and lowest water level was approximately -4 inches. The licensee reported that the 2C battery charger was back in service but remained in the LCO due to battery requirements. They are also completing engineering evaluations for extent of condition. The NRC Resident Inspector was notified. R1DO (Noggle) notified.

• * * UPDATE PROVIDED BY LICENSEE (KLINEFELTER) TO NRC (HELD) AT 1710 ON 4/12/05 * * *

At 16:28 on April 12th, 2005 Susquehanna Unit 2 entered Mode 2 (Start-up) following the completion of repairs to the failed 125VDC battery charger and inspection to the remaining 3 Unit 2 125VDC battery chargers. A courtesy call was made (to) the Pennsylvania Emergency Management Agency. An informational press release will be made. The NRC Resident Inspector was notified by the licensee. R1DO (White) was notified.

At 0731 hours (EDT) on 04/26/2005 the Unit 2 SPDS system was removed from service for planned maintenance. The duration of work is expected to be 48 hours (scheduled for completion at 0700 hours (EDT) on 04/28/2005). ERDS will remain operable during the work window but several points will not be available. For example 23 of 58 ERDS points will be unavailable while SPDS is out of service. However, the ERDS system will still be operable and transmit the remaining points. Loss of Emergency Assessment Capability - A review of the ability of the emergency organization to function without SPDS was performed. Alternate sources for many of the points in SPDS were identified and are contained on an Emergency Plan format in PICSY (plant integrated computer system). Those points not available from PICSY can be obtained from the control room. With these compensatory actions and the communications in place between the facilities, there will not be a major loss of emergency assessment capability. Since the Unit 2 SPDS computer system will be unavailable for greater than 8 hours, this is considered a Loss of Emergency Assessment Capability and reportable under 10CFR50.72(b)(3)(xiii). The NRC Resident Inspector has been notified.

• * * UPDATE ON 4/28/05 AT 1412 EDT FROM RON FRY TO A. COSTA * * *

This is a followup courtesy notification to EN#4160. The Unit 2 SPDS system as noted below (above) has been restored to normal at 13:52 (EDT) on 4/28/05 The licensee notified the NRC Resident Inspector. Notified R1DO (Barkley).

On 10/28/2005, Susquehanna Unit 1 operators began the process of shutting down the unit for a maintenance outage to address control cell friction. It was expected that during this evolution, "slow to settle" control rods would be encountered, and rather than delay the continuance of the shutdown to perform operability testing, it was determined that these control rods would conservatively be declared inoperable. Based on previous testing data and trending analysis of operability testing performed during the cycle, PPL (Pennsylvania Power and Light) believes that the subject control rods would have passed operability tests. Accordingly, at 2332 EDT, Technical Specification 3.1.3.f was entered which requires the unit to be taken to Mode 3 in 12 hours when nine or more control rods are inoperable. Entry into this specification was also required per 3.1.3.e when it was determined that four inoperable rods resided within one Banked Position Withdrawal Sequence (BPWS) group. The inoperable control rods are fully inserted and disarmed, shutdown margin requirements are met, and the control rod system was always fully capable of performing its safety function. At the time when the ninth control rod was declared inoperable, Unit One was operating at 18% power. Unit Two is continuing operation at 100% power. PPL is reporting this event as a Technical Specification required shutdown per 10CFR 50.72(b)(2)(i). The NRC Resident Inspector was notified by the licensee. Unit 2 is not affected by this problem and will continue to operate at 100% power.

• * * UPDATE FROM LICENSEE (ILLIADIS) TO NRC HUFFMAN AT 1011 EDT ON 10/29/05 * * *

At 0744 the plant reached All Rods In and at 0805 entered Mode 3 in accordance with plant procedures to fulfill the requirement of Technical Specification 3.1.3. As stated in the original EN# 42093, it was expected that during this evolution, "slow to settle"' control rods would be encountered, and rather than delay the continuance of the shutdown to perform operability testing, it was determined that these control rods would conservatively be declared inoperable. This action was taken on a total of 41 control rods that failed to settle at position 00 during the shutdown. There are 4 additional control rods that were declared inoperable previously during the cycle and subsequently settled to 00. Of the 41 control rods that failed to settle at the time of the shutdown, 6 have settled to 00 at the time of this update. Based on previous testing data and trending analysis of operability testing performed during the cycle, PPL continues to believe that the subject control rods would have passed operability tests. The inoperable control rods are fully inserted, shutdown margin requirements are met, and the control rod system was always fully capable of performing its safety function. There were no ECCS initiations and no challenges to containment. The licensee notified the NRC Resident Inspector. R1DO (Silk) notified.