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	Q	Significance	CCA	Identified by	Title	Description
05000277/FIN-2000013-01	2000Q4	Severity level IV		NRC identified	N/A	The team identified a non-cited Severity Level IV violation of 10 CFR 55.31(a)(4) because an operator license application was submitted to the NRC in August 1999 with incorrect information. The application was incorrect because it indicated that the individual completed all required training even though the emergency preparedness portion of his required training was not completed until May 2000 (approximately eight months after the individual had been licensed) When evaluating this issue according to NRC Manual Chapter 0610*, Appendix B, it did involve extenuating circumstances in that the issue potentially impacted the NRCs ability to perform its regulatory function. The teams evaluation of the apparent cause indicated a problem between the emergency preparedness and operator training organizations, and limited to one individual. The issue was documented in PECOs corrective action program as Performance Enhancement Program Issue I0012084. (Section 4OA2.a)
05000277/FIN-2003004-04	2003Q3	Severity level IV		NRC identified	Inadequate Emergency Plan Change Documentation, 10 CFR 50.54(Q)	The inspector identified a Severity Level IV non-cited violation of 10 CFR 50.54(q). During the implementation of a new Standard Emergency Plan, Exelon did not retain a record that determined whether a decrease-in-effectiveness had or had not occurred when Exelon generated the new Standard Emergency Plan that deleted portions of the previous Combined Limerick/Peach Bottom Emergency Plan. Changing emergency plan commitments without documentation impacts the NRC's ability to perform its regulatory function and is, therefore, processed through traditional enforcement as specified in Section IV.A.3 of the Enforcement Policy, issued May 1, 2000 (65 CFR 25388). According to Supplement VIII of the Enforcement Policy, this finding was determined to be a Severity Level IV because it involved a failure to meet a requirement not directly related to assessment and notification.
05000277/FIN-2003008-01	2003Q1	Severity level IV		NRC identified	10 CFR 50.54(Q) Violation for Decreasing the Effectiveness of the Plan by Changing Eals That Address Toxic Gas Without Prior NRC Approval	Severity Level IV. The licensee changed its emergency action level schemes such that there would be a reduction in declarable events as the emphasis shifted from personnel safety to equipment status. The changes were determined to be a decrease in the effectiveness of the emergency plans. Decreases in the effectiveness of an emergency plan must Page 7 of 8 receive NRC review prior to implementation. The changes were implemented without NRC approval. The finding was determined to be more than minor as its significance was related to the impact it would have on the mobilization of the emergency response organization and preclude offsite agencies from being aware of adverse conditions on site. The licensee accepted the NRC's position and entered this issue into its corrective action program (Condition Report 139997) and will change the emergency action levels back to the original wording. The implementation of the changes which decreased the effectiveness of the emergency plans, without NRC review, is being treated as a non-cited violations consistent with Section VI.A of the Enforcement Policy, issued on May 1, 2000 (65 FR 25388).
05000277/FIN-2003013-01	2003Q4	White		Self-revealing	Failure to Adequately Maintain the E2 Emergency Diesel Generator	A self-revealing finding was identified for the failure to adequately maintain the E2 emergency diesel generator (EDG) between July 1992 and September 2003. This finding involved two apparent violations. An apparent violation of Technical Specifications was identified for the failure to maintain the maintenance procedure for installation of EDG adapter gaskets. The procedure did not incorporate certain vendor recommendations intended to provide proper sealing of the gaskets, leading to relaxation over several years that allowed combustion gases to enter the jacket coolant system. An apparent violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions was identified because Exelon did not correct a condition adverse to quality following two instances of low jacket water pressure observed on the E2 emergency diesel generator (EDG) in March and April 2003. Subsequently, the EDG failed due to a low jacket water pressure condition. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The finding was assessed using a Phase 3 evaluation. The finding is of low to moderate safety significance (WHITE) at Unit 2 based on delta core damage frequency ( CDF) and delta large early release frequency ( LERF). The finding is of very low safety significance (GREEN) at Unit 3 based on CDF and LERF. The difference between the two units is attributable to differences in electrical bus loads. (Section 2.2)
05000277/FIN-2004003-03	2004Q2	Severity level IV		Licensee-identified	Licensee-Identified Violation	10 CFR 55.25 requires in part, that the facility licensee notify the Commission within 30 days of discovery, that a licensed operator has been diagnosed with a permanent physical condition that adversely affects the performance of assigned operator job duties, so that the Commission can make a determination of the licensed operators medical fitness. Contrary to this requirement on March 20, 2003, the facility licensee identified that a licensed operator underwent a medical procedure in December 1998 that should have been reported to the NRC. This issue was of very low safety significance because upon review of additional information provided by the facility licensee, the NRC physician determined that a restriction would not have been required because the licensed operator would have been able to perform licensed responsibilities without impairment. This failure to report medical information to the NRC impacted the regulatory process, and therefore, is classified at Severity Level IV.
05000277/FIN-2007002-04	2007Q1			NRC identified	Incorrect size breaker resulted in a fire in the 4T4 480 volt load center	At approximately 9:16 a.m. on February 27, 2007, a fire was suspected to have started based on the receipt of numerous secondary plant alarms in the main control room (MCR) and the report of smoke near the 4T4\' 480 Volt load center. The inspectors responded to the MCR following a site announcement for the fire brigade to respond to a suspected fire in the Unit 3 turbine building. The inspectors monitored the operators response to the event and the status of plant equipment. The observations were primarily focused on the nuclear safety aspects of the plants and operators responses. The inspectors also monitored the response of PBAPSs emergency response organization to the declaration of an UE. Subsequent to the fire, the inspectors discussed the fire with operations, engineering and PBAPS management personnel to gain an understanding of the event and to assess their followup actions. The inspectors reviewed operator logs and operators actions taken in accordance with licensee procedures. Based on the operators narrative logs, the fire brigade was dispatched to the Unit 3 turbine building at approximately 9:20 a.m. Fire personnel investigated and notified the MCR that an actual fire existed at 9:38 a.m. An Unusual Event for a fire not extinguished within 15 minutes (emergency action level (EAL) HU6) was declared at 9:41 a.m. All state and local government notifications were completed by 9:59 a.m. and the NRC Headquarters Operations Officer was notified of the event at 10:36 a.m. The fire was considered to be extinguished at approximately 10:32 a.m. At 11:37 a.m., the Unusual Event was terminated. Prior to the report of the potential fire, Unit 3 was operating at full power. As a result of fire and the associated response actions, numerous non-safety-related loads powered by the 4T4\' 480 Volt load center were de-energized. Equipment that was de-energized included: the B isophase bus cooler fan, the B drywell chiller, the B recirculation pump speed controller, the leading edge flow meters and the B reactor feed pump. Plant operators took the required TS actions and responded to the equipment losses by performing controlled reactor power reductions and stabilized the plant at approximately 50 percent of rated power. The inspectors verified that the required reports were made during the event and that no further reports are planned. The inspectors also verified that this issue (IR 569889) was placed into the CAP. Preliminarily, PBAPS has determined that the fire resulted from an apparent mismatch between the ratings of one breaker and its cubicle in the 4T4\' 480 volt load center. A root cause investigation was ongoing at the end of the inspection period and will be reviewed by the inspectors during a future inspection period. At the close of this inspection period, the inspectors were reviewing the event and awaiting the results of the root cause evaluation to understand the potential performance deficiencies. This issue is unresolved pending review of PBAPSs causal evaluation and corrective actions by the inspectors to characterize the issue. URI 05000277/2007002-04, Incorrect Size Breaker Resulted in a Fire in the 4T4\' 480 Volt Load Center.
05000277/FIN-2007002-05	2007Q1			NRC identified	Missed procedure step resulted in unplanned overloading of the E-3 EDG	The inspectors reviewed selected applicable plant records, correction action documents and approved procedures while evaluating the performance of operations personnel in response to non-routine evolutions. The inspectors assessed personnel performance to determine what occurred and how the operators responded, and to determine if plant personnels response was in accordance with plant procedures and training. The following non-routine evolution was reviewed: During the conduct of surveillance testing of the E-3 EDG on March 15, 2007, a licensed operator missed the performance of a required step in a supporting system operating procedure. The omission of the procedure step placed the E-3 EDG in the isochronous mode while synchronized with offsite power through a 4 kilovolt (kV) vital bus. This condition resulted in unexpectedly loading the E-3 EDG beyond its 30-minute load rating. The ST and supporting procedures directed the synchronization of the E-3 EDG to a selected 4 kV bus to pick up the bus loads. The procedure subsequently directed opening the offsite power feeder breaker to the 4 kV vital bus (the missed step) before placing the EDG in the isochronous mode. PBAPS placed this issue in the CAP by initiating IR 604364. Prompt corrective actions included the selected implementation of additional peer checking of procedure performance place-keeping. The E-3 EDG was inspected for potential damage and tested before being returned to an operable condition in accordance with TS on March 17, 2007. The causal evaluation of this event was ongoing at the end of the inspection period. At the close of this inspection period, the inspectors were reviewing the event and awaiting the results of the causal evaluation to understand the potential performance deficiencies. This issue is unresolved pending review of PBAPSs causal evaluation and corrective actions by the inspectors to characterize the issue. URI 05000277/2007002-05, Missed Procedure Step Resulted in Unplanned Overloading of the E-3 EDG.
05000277/FIN-2007405-01	2007Q4	White		NRC identified	Security	Inattentive security officers and the staff determination that the licensee failed to effectively implement its behavior observation program
05000277/FIN-2008002-01	2008Q1	Green	P.3	NRC identified	Failure to Identify and Document Fire Brigade Deficiencies (Section 1R05.2)	The inspectors identified a non-cited violation (NCV) of Technical Specification (TS) 5.4.1, which requires that written procedures be implemented covering the Fire Protection Program. The Fire Drill Performance procedure was inadequately implemented because numerous fire brigade deficiencies were not discussed at the post-drill critique or documented in the fire drill record. The licensee has entered this problem into their CAP for action. This finding is more than minor because it affects the impairment or degradation of a fire protection feature, specifically, on the ability of the fire brigade to effectively carry out the defense-in-depth attribute of manual fire fighting and is associated with the Mitigating Systems Cornerstone and its respective attribute of human performance. This finding is of very low safety significance because the observed crew was only one of five crews of the site fire brigade team, the other crews had no known problems, and the overall condition of the fire detection and suppression systems had been satisfactory. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution because Peach Bottom Atomic Power Station personnel did not properly identify and assess deficiencies with the fire brigades performance. (IMC 0305, aspect P.3 (a)) (Section 1R05.2).
05000277/FIN-2008003-01	2008Q2	Green		Self-revealing	Foreign Material Discovered in Fire Valve	A self-revealing NCV of PBAPSs Unit 2 License Condition 2.C (4), Fire Protection Program, was identified when maintenance personnel discovered foreign material inside a supply valve to an automatic 13KV switchgear sprinkler system installed because there is a one-hour rated raceway encapsulated in the 13KV switchgear area. The Fire Protection Program requires automatic suppression when one-hour rated raceway encapsulation is used. PBAPS has removed the foreign material, replaced the affected valve, and entered this issue into their CAP for appropriate action. The inspectors determined that there was no cross-cutting aspect to this finding. The finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of protection against external factors (i.e., fire), and it affects the objective of ensuring reliability and capability of systems that respond to initiating events. The finding was of very low significance because PBAPS demonstrated that the core damage frequency (CDF) associated with a fire in this area was in the 1 E-7 range for all assumed fires. (Section 1R12
05000277/FIN-2008004-01	2008Q3	Green	P.2	Self-revealing	Inadequate EOC Review Results in Delay in Discovery of ESW Leaks	A self-revealing NCV of 10 CFR Part 50, Appendix B, Criteria V, Instructions, Procedures and Drawings, was identified for a failure to follow procedure, WC-PB-2000, Outage Control Center (OCC) Emergent Issue Response, that resulted in an inadequate extend of condition (EOC) evaluation being performed for an emergency service water (ESW) leak that was discovered on the E-1 emergency diesel generator (EDG). Specifically, Operations personnel failed to look at similar ESW locations on the E-2, E-3, and E-4 EDGs. This resulted in a nine-day delay in discovering a similar leak on the E-4 EDG. This finding is greater than minor because it is similar to the example 4a., Insignificant Procedural Errors, in Manual Chapter 0612, Appendix E, in that, the later evaluation of the ESW leak discovered on the E-4 EDG resulted in safety-related equipment being adversely affected. Using the Phase 1 worksheet in Manual Chapter 0609, Significance Determination Process, the finding was of very low safety significance (Green) since it did not represent an actual loss of system safety function for the ESW system. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution (PI&R), Corrective Action Program (CAP) because the licensee failed to thoroughly evaluate the EOC of the leak on the E-1 EDG and it resulted in a nine-day delay in discovering additional leaks associated with the E-3 and E-4 EDGs. (IMC 0305, aspect P.1(c)). (Section 1R15)
05000277/FIN-2008004-02	2008Q3	Green	P.6 P.3(c)	NRC identified	Failure to Comply with 10 CFR 20 Appendix G	The inspectors identified a NCV of 10 CFR 20, Appendix G, Section III.C. 5. Specifically, the licensee did not conduct a Quality Assurance Program sufficient to assure conformance with 10 CFR 61.55, in that, the program was not adequate to identify incorrect gamma spectroscopy analyses of a principal gamma emitting radionuclide used to scale hard-to detect radionuclides for purposes of waste classification in accordance with 10 CFR 61.55. The licensee entered the deficiency into its CAP (IR799894). The failure to conduct a sufficiently robust 10 CFR 61 Quality Assurance Program, to assure conformance with 10 CFR 61.55, is a performance deficiency that was reasonably within the licensees ability to foresee and correct, and which should have been prevented. The finding is more than minor because it affected the associated cornerstone objective in that the licensees 10 CFR 61 Quality Assurance Program did not identify incorrectly analyzed waste samples used to classify radioactive waste for land disposal. This finding was determined to be of very low safety significance because no radiation limits were exceeded, there was no breach of packaging, there was no certificate of compliance finding, there was no low level burial ground non-conformance, and lastly, there was no failure to make notifications or provide emergency notification information. The cause of this finding was related to the cross-cutting area of PI&R, self and independent assessments component, in that, although actions were taken to coordinate and communicate results from assessments to affected personnel, corrective actions were not sufficiently comprehensive to identify incorrect vendor analyses (IMC 0305, aspect P.3(c)). (Section 2PS2)
05000277/FIN-2008004-03	2008Q3			NRC identified	Failure to Make a 10 CFR 50.72(b)(2)(xi) Notification	The inspectors identified a NCV of 10 CFR 50.72(b)(2)(xi) because the NRC Operations Center was not notified of a reportable event. Specifically, PBAPS did not formally report, to the NRC Operations Center, a planned press release and the notification of other government agencies regarding a transformer fire and petroleum product spill event that occurred on July 23 and 24, 2008. This deficiency was evaluated using the traditional enforcement process since the failure to make a required report could adversely impact the NRCs ability to carry out its regulatory mission. This event was related to pubic health and safety because it involved a fire emergency that contributed to the loss of two of the plants three offsite power sources. This event was related to protection of the environment because it involved the spill of a more than minor quantity of oil that required reporting to the State of Pennsylvania. While reviewing this finding, the inspectors considered the fact that the NRC was informally notified. The inspectors considered the above and evaluated the severity of this violation using the criteria contained in Supplement I Reactor Operations and Section VI.A.1 of the NRCs Enforcement Policy and determined that this finding met the criteria for disposition as a NCV. (Section 4OA3.1
05000277/FIN-2008005-01	2008Q4	Green	H.4 H.5	Self-revealing	Incorrect Performance of Procedure Step Resulted in Inoperability of a DC Bus for Longer than the TS Allowed Outage Time	A self-revealing (Green) NCV of Technical Specification (TS) 5.4.1 was identified when operators inadequately implemented an abnormal operating (AO) procedure on two occasions. Specifically, an event where the Unit 2 Division II direct current (DC) electrical power subsystem was inoperable for longer than the allowed outage time specified in Unit 3 TS 3.8.4, resulted from PBAPS personnel not recognizing the existence of conflicting procedure guidance and the improper removal of a configuration control tool. This finding is more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, and impacted the cornerstone objective of ensuring the reliability of the Unit 2, Division II, DC electrical power subsystem to respond to initiating events, in that, one of its associated battery chargers was being supplied from a non-qualified alternating current (AC) power source. The inspectors concluded that this finding affected the Mitigating Systems Cornerstone and answered \"No\" to all relevant questions. Specifically, the supply of a non-qualified AC power source to the Unit 2, Division II DC electrical power system was a qualification issue confirmed not to result in a loss of functionality. Although the Unit 2, Division II DC electrical power system was inoperable for longer than its 12-hour TS allowed outage time, this qualification issue did not result in an actual loss of safety function. Therefore, this finding was considered to be of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of human performance (work control component) because PBAPS personnel did not adequately coordinate work activities by incorporating actions to address: the impact of changes to the work scope or activity on the plant and human performance; nor the need to keep personnel apprised of the operational impact of work activities; and plant conditions that may affect work when conflicting procedures led to inadequate procedure adherence and the unplanned inoperability of the Unit 2 Division II DC electrical subsystem. (IMC 0305 aspect: H.3(b)). (Section 4OA3.1
05000277/FIN-2008007-01	2008Q2	Green	P.3	NRC identified	Inadequate Battery Connection Resistance Testing (Section 1R21.2.1.1)	The team identified a finding of very low safety significance involving a non-cited violation of Technical Specification (TS) 3.8.4.5, in that Exelon did not verify certain battery connection resistances were within the TS limit. Specifically, Exelon did not verify the inter-tier connection resistances to be within the TS 3.8.4.5 limit of less than or equal to 40 micro-ohms every 12 months. The team determined that Exelon exempted the inter-tier connections from the testing requirement. In response, Exelon performed the required testing and identified a connection in the 2B battery that was greater than the TS limit. Exelon restored the degraded connection and initiated actions to revise the surveillance test procedures to incorporate all battery connections. This issue was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because it did not result in a loss of safety system function. Because the licensee had previously identified a similar inadequacy in the test procedure, the finding had a cross-cutting aspect in the area of Problem Identification and Resolution - Corrective Actions. (IMC 0305, aspect P.1(d)) (1R21.2.1.1)
05000277/FIN-2008007-02	2008Q2	Green	H.7	NRC identified	Non-conservative Inputs Used in Design Calculations for Offsite Power Operability (Section 1R21.2.1.4)	The team identified a finding of very low safety significance, in that Exelon failed to use appropriate inputs in design calculations as required by Exelon Procedure CC-AA-102 - Design Input and Configuration Change Impact Screening. The requirements of the procedure include ensuring performance requirements are the maximum or minimum numerical values of specific design parameters, specifically, the Maximum time to automatically initiate a system action. The team determined the response speed used by Exelon for the automatic load tap changer (LTC) controller and mechanism for the stations startup transformers, in the calculation to determine offsite power availability, was non-conservative. This assumption resulted in the grid voltage limit, used to assess technical specification offsite power supply operability, to be nonconservative. In response, Exelon performed preliminary calculations with revised LTC times, which showed that the offsite grid remained operable at the specified voltage limits. Exelon entered the issue into the corrective action program to re-perform the calculation and raise the allowed offsite grid voltage level. This finding was more than minor because it is associated with the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because it was design deficiency that did not result in a loss of offsite power operability. Because the licensee had recently performed this calculation with the non-conservative inputs, the finding has a cross-cutting aspect in the area of Human Performance - Resources. (IMC 0305, aspect H.2.(c)) (1R21.2.1.4)
05000277/FIN-2008007-03	2008Q2			NRC identified	Vital Bus Degraded Voltage Protection (Section 1R21.2.1.8)	The team reviewed Exelons load flow and vital bus voltage calculations. The review was performed to verify the minimum vital bus voltage needed to ensure operation of safety related loads required during design basis events was adequate. The team determined that voltages used in these analyses were not based on the trip set point of the Technical Specification Function 4 (LOCA) degraded voltage relay. Exelon had used voltages higher than were afforded by the Function 4 relays based on their belief that minimum expected value, as defined in GL 79-36 Adequacy of Station Electric Distribution Systems Voltage, could be used to calculate adequate voltages to vital loads. In using this assumption Exelon credited voltage improvement due to operation of the non-safety related startup transformer load tap changers in their analysis. The tap changer restores voltage to the vital bus during and following the post accident voltage transient. The team reviewed NRC Letter dated June 2, 1977, which was the basis for the licensing requirement to install the degraded voltage relay protection scheme. This licensing requirement required the set points for the second level reduced-voltage relays provide adequate voltage, from offsite or onsite power sources, for safety related loads at all onsite system distribution levels. The inspectors reviewed the PBAPS licensing records related to degraded voltage protection and did not find where the NRC had allowed Peach Bottom to credit operation of automatic tap changers in lieu of the technical specification reduced voltage relays to provide protection. Exelon stated that their approach was acceptable and the NRC had given this credit when it reviewed and approved certain voltage studies submitted as part of licensing actions related to the degraded voltage relays. This unresolved issue is being opened to determine if the Peach Bottom approved licensing bases includes the use of automatic load tap changers to protect the vital bus from unacceptable low voltage conditions during loss of coolant accidents. (URI 05000277;278/2008007-003, Vital Bus Degraded Voltage Protection)
05000277/FIN-2008405-01	2008Q1	Severity level IV		NRC identified	Extent of Condition and Corrective Action Program Usage for Operator Watch Standing Issues. (Section 4OA2.2)	On September 10, 2007, representatives of WCBS-TV (New York City) contacted the NRC stating that they possessed videotapes of inattentive security officers at the Peach Bottom Atomic Power Station (PBAPS). Based upon this information, the NRC Region I Regional Administrator directed implementation of enhanced inspection oversight of security activities by the resident inspectors at PBAPS, and verbally informed Exelon management of the information received. Exelon commenced an internal investigation based upon this information. On September 19, 2007, WCBS-TV shared the videotapes with the NRC staff, which viewed the videos and determined that the situation warranted an Augmented Inspection. An Augmented Inspection Team (AIT) completed an inspection at PBAPS from September 21 through 28, 2007. The team concluded that Exelons prompt compensatory measures and corrective actions in response to the videotaped inattentive security officers at PBAPS were appropriate and ensured the stations ability to satisfy the Security Plan. However, the team determined that the security officer inattentiveness affected the defense-in-depth strategy, and that security force supervisors were not effective in ensuring unacceptable behavior was promptly identified and corrected. The AIT inspection results were published on November 5, 2007 in NRC Inspection Report 2007404 (ADAMS accession number ML073090061). On October 4, 2007, Exelon sent a letter to the NRC Region I Regional Administrator (ML072850708) which described their completed actions and initiatives to address the issues identified by the AIT. These initiatives included terminating the current security contract with their contractor and transitioning to a proprietary security force. Exelon also described plans to complete a root cause analysis of the security officer inattentiveness, identify corrective actions, and perform safety conscious work environment (SCWE) surveys of the Peach Bottom Security organization. On October 19, 2007, the NRC issued a Confirmatory Action Letter (CAL) to confirm Exelons commitments to assure that security officers remain attentive at all times while on duty (ML072920283). Exelon completed their root cause analysis in October 2007 and identified several causal factors related to the security officer inattentiveness issues and specific corrective actions to address the causal factors. One of the corrective actions was to perform a systematic SCWE assessment of all work groups at PBAPS (including the Security work group) based on an integrated review of information from the PBAPS Corrective Action Program (CAP), Employee Concerns Program (ECP), publicly available NRC allegation statistics, and SCWE surveys. The NRC conducted an AIT follow-up inspection from November 5 through 9, 2007, to review Exelons root cause analysis report and their planned corrective actions. The inspectors concluded the corrective actions were appropriate. With regard to the security officer inattentiveness issue, the AIT follow-up inspection identified a finding regarding Exelons failure to maintain the minimum required number of available security officer responders and an associated failure to implement an effective behavior observation program. The AIT follow-up inspection determined that the finding was related to SCWE because it involved security supervisors who did not encourage the free flow of information related to raising safety concerns, and who did not respond to security officer safety concerns in an open, honest, and non-defensive manner. The NRC determined the finding was of low to moderate safety significance (White). This was documented in a subsequent letter to Exelon dated February 12, 2008 (ML080440012). The AIT follow-up inspection results were issued in NRC Inspection Report 2007405 (ML073550590) dated December 21, 2007. Region I determined that Exelons actions to address the PBAPS inattentive security officer issues and their plans to transition to a proprietary security force warranted additional inspection and oversight beyond that specified in the Reactor Oversight Process (ROP) baseline inspection program. On November 28, 2007, the Regional Administrator recommended, through a Deviation Memorandum to the NRCs Executive Director for Operations (EDO), that PBAPS warranted additional inspection resources (ML073320344). One additional inspection activity was to conduct inspections of Exelons efforts to address SCWE issues, including a review of the results of SCWE surveys conducted at the site. The EDO approved this request on November 28, 2007. Consistent with the planned corrective actions from their root cause evaluation, Exelon arranged for a third party to conduct a survey of the SCWE at PBAPS. The survey was in the form of a series of questions provided to the staff in January 2008. The survey was completed and the results provided to Exelon in February 2008. A separate SCWE survey of the security organization was also conducted during November 2007. Exelon utilized the survey results to complete a self-assessment of the SCWE at PBAPS. In accordance with the NRC Action Matrix Deviation Memorandum, this inspection was conducted onsite from March 24 though 28, 2008, to review Exelons self-assessment of the PBAPS SCWE, including a review of the results of their SCWE survey. Other completed Deviation Memorandum activities included a security organization performance monitoring inspection (ML080720038) and a root cause corrective action evaluation (ML081090161).
05000277/FIN-2009002-01	2009Q1	Green	H.5	Self-revealing	Inadequate Work Instructions Result in Inadvertent ESF Actuation	A self-revealing NCV of 10 CFR 50 Appendix B, Criteria V, Instructions, Procedures and Drawings was identified when inadequate work instructions resulted in a momentary shorting of a terminal lead during maintenance, which caused an inadvertent Unit 3, primary containment isolation valve (PCIV) signal and entry into a one-hour shutdown Technical Specification (TS) Action Statement on March 3, 2009.Specifically, the work instructions allowed the technicians to lift and manipulate energized leads on a safety-related pressure switch without providing any guidance as to the risk and consequences that inadvertent grounding of those energized leads could cause. Because the risk and consequences were not considered and an inadvertent grounding occurred, a PCIV signal resulted that closed normally open valves on both the containment atmosphere control (CAC) system and the instrument nitrogen system containment penetrations. In addition, both PCIV valves on containment atmosphere dilution (CAD) system were rendered inoperable which required the operators to enter an unplanned one-hour TS Action Statement(3.6.1.3.B) and would have required a plant shutdown within the following 12 hours. Corrective actions included replacing the blown fuse, entering the issue into the CAP, and making a required 60 day verbal report to the NRC. The finding is more than minor because it could reasonably be viewed as a precursor to a significant event. Specifically, the failure to assess the risk of inadvertent grounding of energized leads on safety equipment could pose a credible hazard as an initiating event during plant operation. The finding was of very low safety significance because the valves in question failed closed and did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the area of human performance (work control) because the licensees work instructions did not provide appropriate risk insights regarding the risks associated with potential grounding of the energized leads. H.3(a
05000277/FIN-2009002-02	2009Q1	Green	H.7	Self-revealing	Inoperable \'A\' WRNM Results in a Condition Prohibited by TSs	A self-revealing, Green NCV of Unit 3 TS 3.0.4 was identified by the inspectors on January 26, 2009, when a half-scram occurred on Unit 3, shortly after Unit 3 entered Mode 2 for plant startup. Specifically, the A Wide-Range Neutron Monitoring (WRNM) was inoperable as a result of inadequate procedural guidance regarding adjustments made to the mean square voltage (MSV) offset during the outage (prior to the January 26, 2009, startup). The inadequate procedural guidance allowed adjustments to be made which resulted in the WRNM not making a smooth transition from the counting region to the MSV region of operation, causing the AWRNM to be inoperable and resulting in an unexpected half-scram when the WRNM transitioned from the counting region to the MSV region of operation. As a result, TS3.3.1.1 requirements for the number of available channels of WRNM short period RPS trip in Mode 2 had not been met. TS 3.0.4 requires that when a LCO is not met, entry into a mode or other specified condition shall only be made when the associated actions to be entered permit continued operation in the mode or other condition specified for an unlimited period of time. Corrective actions included entering the issue into the CAP, conducting an event review, and submitting a License Event Report (LER) to the NRC, and revising the WRNM adjustment procedure. The finding is more than minor because it is associated with the procedure quality attribute and adversely affected the Initiating Events Cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. The finding was of very low safety significance because it did not contribute to the likelihood that both a reactor trip would occur and that mitigation equipment would not be available. This finding has a cross-cutting aspect in the area of human performance (resources) because the licensees procedure did not provide adequate guidance to prevent adjusting the MSV offset to an unacceptable value. IMC 0305 aspect: H.2(c
05000277/FIN-2009002-03	2009Q1	Green		NRC identified	Departure from a Method of Evaluation without Prior NRC Approval	An inspector-identified, Severity Level IV NCV of 10 CFR 50.59 was identified when PBAPS made a safety analyses change that departed from a method of evaluation described in the UFSAR without obtaining prior NRC approval and a license amendment. Specifically, PBAPS used a spent fuel pool criticality analysis methodology that was not previously approved by the NRC, and did not adopt an NRC-approved methodology en toto and apply it consistent with applicable terms, conditions, and limitations of that methodology. Corrective actions for this problem included entering the issue into the CAP and making plans to develop a technical evaluation that would demonstrate, using methodologies approved for PBAPS, that adequate margin to criticality exists for the nonconforming condition presented by degraded Boraflex in the SFP storage racks. Additionally, PBAPS submitted a LAR, to use alternative SFP criticality analyses, to the NRC on June 25, 2008. This deficiency was evaluated using the traditional enforcement process since it potentially impacts or impedes the NRCs ability to carry out its regulatory mission, in that, PBAPS did not request and receive prior NRC approval for changes in licensed activities. The finding is more than minor and a Severity Level IV violation because it is similar to example D.5 of Supplement I, Reactor Operations, to the NRCs Enforcement Policy. Specifically, the finding involved a violation of 10 CFR 50.59that resulted in conditions evaluated as having very low safety significance (i.e., Green) by the SDP. Using the Phase 1 SDP, the inspectors determined that the condition resulting from the violation of 10 CFR 50.59 screened to Green because it could affect the functionality of the fuel barrier (cladding)
05000277/FIN-2009002-04	2009Q1			NRC identified	HPCI System Torus Suction Valve Failures	The inspectors identified an unresolved item (URI) related to the adequacy of preventive maintenance on MOVs. On March 12 and 21, 2009, HPCI torus suction valves in Unit 2 and Unit 3, respectively, failed to stroke fully open during routine testing. Dry and hardened stem lubricant was identified in both instances. This issue will remain unresolved pending completion of PBAPSs root cause determination and completion of extent of cause and condition evaluations of MOVs in other accident mitigation systems. On March 12, the Unit 2 HPCI system suppression pool suction valve, MO-2-23-058, failed to fully open when repositioned during quarterly surveillance testing. The valve stroke was interrupted by operation of the motor operator torque switch. On March 21, the Unit 3 HPCI system suppression pool suction valve, MO-3-23-057, failed to fully open when it was repositioned during quarterly testing. The valve stroke was interrupted by actuation of the motor operator torque switch. In both instances, the stem lubricant was found to be dry and hardened. Failures to stroke appeared to be repeat occurrences of a valve failure to stoke event which occurred in October 2007.PBAPS determined that other safety-related MOVs may be similarly affected by the stem lubricant hardening issue. The EOC and extent of cause evaluations were ongoing at the end of the inspection period. These evaluations included selecting a sample of MOVs to be visually examined for dry and/or hardened stem lubricant. In addition, PBAPS selected a number of MOVs for diagnostic testing with monitoring equipment connected to determine if any degradation of MOV capability had occurred since the last diagnostic testing of that MOV. At the end of the inspection period, these activities were still in progress; therefore, this item remains unresolved: URI 05000277, 278/2009002-04, High Pressure Coolant Injection (HPCI) System Torus Suction Valve Failures
05000277/FIN-2009003-01	2009Q2	Green	P.2	Self-revealing	MOV Program Procedures were Inadequate with Regard to Periodicity of Preventitive Maintenance Activities for Stem Lubrication	A self-revealing NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified. Specifically, Exelons Motor Operated Valve (MOV) Program procedures lacked specific instructions to prescribe an acceptable frequency for performing valve stem lubrication, which resulted in test failures of safety related MOVs and affected the reliability of the MOVs safety functions. On Unit 2, the inspectors determined that the finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609, Attachment 4, the inspectors determined that the finding was of very low safety significance (Green)because it was not a design or qualification deficiency, did not represent a loss of system safety function, and was not associated with any external events. On Unit 3, the inspectors determined that the finding was more than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (e.g., containment) protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609, Attachment 4, the inspectors determined that the finding was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment. For both units, this finding has a cross-cutting aspect in the area of Problem Identification and Resolution (PI&R), Corrective Action Program, because PBAPS did not thoroughly evaluate problems such that the resolutions addressed the causes and extent of condition P.1(c). Specifically, PBAPS failed to thoroughly evaluate previous conditions of degraded and hardened grease on safety-related valves, such that the extent of the condition was considered and the cause was resolved.
05000277/FIN-2009003-02	2009Q2	Green	H.8	Self-revealing	Inadequate Procedure Adherence Results in Trip of 3 \\\'A\\\' Recirc Pump and Plant Transient	A self-revealing finding was identified when PBAPS personnel incorrectly performed a maintenance procedure for tuning the reactor recirculation pump (RRP) motor generator (MG) set voltage regulator. Specifically, maintenance personnel adjusted a potentiometer in the wrong direction, which resulted in a trip of the RRP and an unplanned plant transient. This finding is more than minor because the finding is associated with the human performance attribute of the Initiating Events Cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, this error resulted in an unplanned plant transient that reduced reactor power from 75 percent to 33 percent. In accordance with IMC 0609, Attachment 4, the inspectors determined this finding to be of very low safety significance (Green) since the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. This finding has a cross-cutting aspect in the area of human performance, Work Practices, because PBAPS did not define and effectively communicate expectations regarding procedural compliance and personnel did not follow procedures H.4(b). Specifically, PBAPS personnel did not follow procedure IC-11-02011instructions for tuning the 3 A RRP MG set voltage regulator.
05000277/FIN-2009003-03	2009Q2	Green		Licensee-identified	Licensee-Identified Violation	TS 3.1.3, Condition C, requires that control rods that are inoperable for reasons other than being stuck shall be fully inserted and disarmed. TS 3.1.3, Condition E, requires the unit to be in Mode 3 within 12 hours if Condition C cannot be met. On February 11, the 10-51 CRD HCU was declared inoperable for the conduct of maintenance and the TS required actions to fully insert and disarm the CRD were met. Following the completion of maintenance on the HCU, an operator erroneously re-armed the CRD HCU DCVs during the modification of a safety tagging clearance that occurred at approximately 5:30 a.m. on February 12.Over 28 hours later and in excess of the 12-hour completion time allowed by TS3.1.3, PBAPS personnel discovered the error and disarmed the CRD for Control Rod 10-51. PBAPS documented this issue in the CAP as IR 880318. Since Control Rod 10-51 remained fully inserted and there was no loss of safety function during the period of non-compliance, this issue is of very low (Green) safety significance. The LER associated with the event was documented in Section 4OA3.3.
05000277/FIN-2009004-01	2009Q3	Green	H.14	NRC identified	Failure to Perform a 50.59 Review Prior to Installing Jumpers on E Wrnm	An inspector-identified, Severity Level IV NCV of 10 CFR 50.59 was identified when PBAPS made temporary alterations to their facility to address a degradedcondition without performing a 50.59 review. Specifically, PBAPS installed a jumper that bypassed the trip feature of the Unit 3 \'E\' wide-range neutron monitoring (WRNM) system instead of using the WRNM bypass switch as is described in their plant\'s Final Safety Analysis Report (FSAR). Exelon entered this issue into their CAP and the jumper was subsequently removed restoring the original system configuration. Because this was a violation of 10 CFR 50.59, it was considered a violation that potentially impeded or impacted the regulatory process; therefore, this violation was dispositioned using the traditional enforcement process. This finding was more than minor because there was a reasonable possibility that the change requiring a 10 CFR 50.59 Safety Evaluation (SE) would require NRC review and approval prior to implementation in accordance with 10 CFR 50.59(c)(2). This possibility is based on the likelihood that a second WRNM could be bypassed, with the bypass switch built into the WRNM system, without resulting in a trip of the associated reactor protection system (RPS). This condition would be contrary to the design of the WRNM and RPS, thereby . creating the possibility for a malfunction of a structure, system, and component (SSC) important to safety with a different result than any previously evaluated in the FSAR (as updated). Although the SOP is not designed to assess traditional enforcement violations, the NRC assesses the significance of 10 CFR 50.59 violations through the SOP for risk insights. Accordingly, the inspectors evaluated the finding in accordance with IMC 0609, SOP, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems cornerstone. The issue, associated with the installation of the one jumper, was determined to be of very low safety significance (Green) since the issue was determined to be a qualification issue confirmed not to result in loss of operability of the system. This violation involved a facility change that likely would have required a license amendment before its implementation. Comparing this item to the examples in NRC Enforcement Policy, Supplement I, Reactor Operations, this finding is similar to Item 0.5, Violations of 10 CFR 50.59 that result in conditions evaluated as having very row safety significance (i.e., Green) by the SOP. This is a Severity Level IV violation. Additionally, this finding was determined to have a crosscutting aspect in the area of Human Performance, Decision Making component, which states the licensee should use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed action is safe. Specifically, Exelon did not perform a 10 CFR 50.59 safety evaluation or screening when making a temporary alteration to the RPS system which would be installed for the remainder of the operating cycle
05000277/FIN-2009004-02	2009Q3	Green	H.11 H.12	Self-revealing	Inadequate Procedure Adherence Results in the Loss of Safety Function of Systems Supplied by the Sgig System	A self-revealing Green NCV was identified for failure to comply with Technical Specification (TS) 5.4.1 J Procedures, which required that procedures be established, implemented, and maintained for the safety grade instrument gas (SGIG) system. Specifically, the SGIG Pressure Building Circuit Outlet Block Valve (HV-0-7C-1 0) was manipulated without procedure guidance, was out of its normal position, and resulted in the inoperability of certain valves associated with the primary containment and containment atmosphere dilution (CAD) systems for both units. Based on the above, the inspectors determined that manipulating the SGIG Pressure Building Circuit Outlet Block Valve (HV-0-7C-10) without procedure guidance was a performance deficiency that was reasonably within PBAPS\'s ability to foresee and prevent. The inspectors concluded that the manipulating HV-0-7C-10 without a procedure was a more than minor finding because it was associated SSC and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that the containment would protect the public from radionuclide releases caused by accidents or events. Specifically, certain valves associated with the primary containment and containment atmosphere dilution (CAD) systems could not be operated as designed due to this valve being out of its normal position. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC\'s regulatory function, and the finding was not the result of any willful violation of NRC requirements. Accordingly, the inspectors assessed the finding in accordance with IMC 0609, SOP, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Containment Barrier cornerstone. The finding was determined to be of very low safety significance (Green) since the finding did not represent an actual open pathway in the physical integrity of the reactor containment (isolation valves). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Work Practices component, because human error prevention techniques, such as peer and self checking, were inadequately used to prevent mispositioning the SGIG Pressure Building Circuit Outlet Block Valve (HV-0-7C-10). (IMC 0305 Aspect H.4(a)
05000277/FIN-2009004-03	2009Q3	Green		Licensee-identified	Licensee-Identified Violation	The following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy for being dispositioned as a NCV: As documented in report section 40A3.2, LER 05000278/2009004-00 reported a condition prohibited by TS which was discovered when engineering personnel determined that the Unit 3 \'B\' HPSW 1ESW ventilation subsystem was rendered inoperable as a result of preventive maintenance performed on April 13, 2009. The TRM Section 3.11, Engineered Safeguards Compartment Cooling and Ventilation, required immediate compliance with the TS Actions for the inoperability of one HPSW subsystem if one HPSW I ESW pump structure ventilation subsystem is inoperable. TS 3.7.1, Condition A, required action to restore one inoperable HPSW subsystem to an operable status within seven days. TS 3.7.1, Condition B, required the plant be in Mode 3 within 12 hours if Condition A is not met. Contrary to the above, between April 13 and July 5, 2009, the Unit 3 \'B\' HPSW I ESW ventilation subsystem was inoperable and TS 3.7.1 was not entered until the inoperability was discovered on July 3, 2009. PBAPS documented this issue in the CAP as IR 938565. The inspectors reviewed the PBAPS Risk-Informed Inspection Notebook Table 2and concluded that the HPSW I ESW pump structure ventilation system was not required to support HPSW and ESW pump core damage mitigation safety functions. A Region I senior reactor analyst verified this conclusion. Therefore, this issue was of very low (Green) safety significance, because of no impact on the safety function for either subsystem of the HPSW or ESW systems
05000277/FIN-2009005-01	2009Q4	Green	P.2	NRC identified	Continuously Submerged Cables design Deficiency	The inspectors identified an NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, because PBAPS has not maintained safety-related power cables (including low voltage cables) in an environment for which they were designed and tested. Specifically, PBAPS did not adequately select and review for suitability of application of materials a 480 volt ac power cable feeding a safety-related motor control center (E424-0-A) that has been in a submerged environment in manhole 35 for an extended period of time and at least since 2002. Additionally, PBAPS personnel did not take actions to properly evaluate and mitigate the effects of long term submergence of these safety-related electrical power cables. The issue was entered into the licensee\'s CAP as IR 1022206.This finding is more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone and the associated cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated in accordance with IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings and was determined to be of very low safety significance because it did not represent an actual loss of safety function nor contribute to external event core damage sequences. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon did not thoroughly evaluate problems such that the resolutions addressed causes including evaluating for operability conditions adverse to quality P.1 (c). Specifically, station personnel did not adequately evaluate the impacts on operability and service life of operating the cables submerged in water for an extended period of time.
05000277/FIN-2009005-02	2009Q4	Green	P.5 P.2(b)	NRC identified	Failure to Follow Procedures and Implement the Exelon Nuclear Cable Condition Monitoring Program for Non-Safety-Related Control and Power Cables within the Scope of the Maintenance Rule	The inspectors identified a finding for the failure to follow the Exelon fleet procedure for cable monitoring (ER-AA-3003) of non-safety-related cables within the scope of the 10 CFR 50.65 (the Maintenance Rule). Specifically, PBAPS had reported to the NRC that they were implementing this procedure for cables within the scope of GL 2007-01; however, actions were not specified to identify or remediate the cause of repetitive flooding and restore the function of the degraded electrical manhole/vault drain systems. PBAPS initiated IR 1016075 to enter the issues associated with this finding into the CAP. This finding was more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone and the associated cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. This finding was evaluated in accordance with IMC0609.04, Phase 1 - Initial Screening and Characterization of Findings and was determined to be of very low safety significance because it did not represent an actual loss of safety function or contribute to external event core damage sequences. This finding had a cross-cutting aspect in the area of PI&R, Operating Experience, because Exelon did not adequately implement and institutionalize industry operating experience through changes to station processes and procedures P.2(b). Specifically, work order instructions were inadequately scoped in that they were limited to manholes with safety-related cables and did not include all manholes with Maintenance Rule power cables contrary to the scope identified in ER-AA-3003 or GL 2007-01
05000277/FIN-2009005-03	2009Q4	Green	H.11 H.12	Self-revealing	Inadequate Verification Practices while Handling Fuel and Fuel Components	A Green self-revealing NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when PBAPS inadequately implemented human performance tools and verification practices for fuel handling and fuel component handling activities, resulting in a dropped fuel channel in the spent fuel pool (SFP) and a mispositioned fuel bundle in the reactor core during the P3R17refueling outage (RFO). The inspectors verified that corrective actions were promptly performed, including an operability evaluation and video inspection of the SFP racks, and reactor engineering evaluation for the mispositioned fuel bundle. Additionally, the issues were entered into the PBAPS CAP. This finding was more than minor because it was associated with the human performance attribute of the Barrier Integrity cornerstone, and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide release cause by accidents or transients. This finding was determined to be of very low safety significance (Green) in accordance with IMC 0609, Appendix M, SDP Using Qualitative Criteria, because evaluations performed by PBAPS, and verified by the inspectors, determined that there was no actual degradation to the physical barrier integrity. This finding had a cross-cutting aspect in the area of Human Performance, Work Practices, because PBAPS management and personnel did not effectively communicate human error prevention techniques commensurate with the risk of the assigned tasks, such that the work activities were performed safely H.4(a). Specifically, PBAPS management and personnel did not adequately reinforce the importance of using human performance tools and verification practices, including self-check (STAR), concurrent verification, and independent verification, prior to performance of activities involving fuel component handling.
05000277/FIN-2009005-04	2009Q4	Severity level IV		Licensee-identified	Licensee-Identified Violation	The Reload 16, Cycle 17, Revision 4, mid-cycle Core Operating Limits Report (COLR)was prepared and approved between November 21 and 26, 2008. This COLR revision was issued for implementation on March 12, 2009, and was submitted to the NRC by a letter from P. B. Cowan to the U.S. NRC, Issuance of Proprietary and Non-Proprietary COLRs, dated October 1,2009. TS 5.6.5.d, COLR, states, in part, the COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC. Contrary to the above, between its issuance on March 12,2009, and its submittal on October 1,2009, the Reload 16, Cycle 17, Revision4, mid-cycle COLR was not provided in a timely manner to the NRC nor upon its issuance. This issue was documented in the CAP as IR 970608. Traditional enforcement applies since this was a violation that potentially impeded or impacted the regulatory process. This was considered a non-cited Severity Level IV violation since the untimely submittal did not have a material impact on licensed activities
05000277/FIN-2009008-01	2009Q3	Green	P.3	NRC identified	Failure to Take Adequate Cas for Grease Applied to DC Contactors	The inspectors identified a non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, for failure to identify and correct a condition adverse to quality. Specifically, in March 2009, Exelon did not take adequate corrective action to address a procedure deficiency and to ensure that grease inappropriately applied to Cutler Hammer direct current (DC) contactor pivot pins, in an unknown number of DC breakers in the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems at Unit 2 and 3, would be identified and removed in a timely manner. Because the grease could harden over time and cause inadequate DC breaker performance, the inspectors determined that this condition, if left uncorrected, could prevent certain Units 2 and 3 HPCI and RCIC system valves from performing their safety-related function. Exelon entered this issue into their corrective action program as issue report (IR) 950438 and IR 950439. The finding affected the Mitigating Systems cornerstone and was determined to be more than minor because the condition, if left uncorrected, could have become a more significant safety concern. By not requiring, by procedure, the removal of all grease from the affected Cutler Hammer DC contactor pivot pins, Exelon did not ensure that all of the potentially affected DC motor-operated valves in the Unit 2 and Unit 3 HPCI and RCIC systems would be available to perform their design functions if called upon. The inspectors evaluated this finding using Phase I of Manual Chapter 0609 and determined the finding to be of very low safety significance (Green) because it was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system or train safety function, and was not potentially risk significant due to external events. This finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because Exelon failed to take appropriate corrective actions to address a safety issue in a timely manner, commensurate with the safety-significance and complexity P.1(d). Specifically, Exelon did not take appropriate corrective actions to ensure that grease inappropriately applied to Cutler Hammer DC contactor pivot pins would be, by procedure, identified and removed in a timely manner
05000277/FIN-2010002-01	2010Q1	Green	P.2	Self-revealing	Inadequate Corrective Action to Address Multiple Slow Control Rods with Adverse SSPV Diaphragms	A self-revealing, Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, occurred when PBAPS failed to identify and correct a condition adverse to the quality. Specifically, an issue related to control rod drive scram solenoid pilot valve (SSPV) diaphragms, as described in vendor documents and NRC generic communication, was not corrected after several slow control rods were identified during scram time testing between 2004 and 2010. Consequently, 21 slow rods were identified during Unit 2 scram time testing that was conducted from January 30 to January 31, 2010. PBAPS immediately performed maintenance to replace the defective SSPV Diagrams on all 21 Unit 2 slow control rods by February 1, 2010, and successfully performed post-maintenance scram time testing. Additionally, the issues were entered into the PBAPS CAP. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems (MS) cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the phase 1 worksheet in Attachment 4 of IMC 0609, Significance Determination Process, the inspectors determined that the finding affected the MS cornerstone and was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of safety system function, and was not associated with any external events. The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification & resolution (PI&R), CAP, because PBAPS did not thoroughly evaluate previously identified conditions adverse to the quality of the SSPV diaphragms, such that the resolution addressed the cause and extent-of-condition
05000277/FIN-2010002-02	2010Q1	Green		Licensee-identified	Licensee-Identified Violation	TS Limiting Condition for Operation 3,6,1,3, Condition A, requires a main steam line flow path to be isolated within eight hours when one MSIV is inoperable in Modes 1, 2, and 3, TS 3,6,1,3, Condition F, requires the unit to be in Mode 3within 12 hours, and Mode 4 within 36 hours, if Condition A cannot be met. Contrary to the above, on September 18, 2009, an engineering evaluation determined that the outboard MSIV AO-3-01A-086A did not meet its required TS minimum closure time of greater or equal to three seconds, This determination was based on MSIV stroke time testing performed on September 14,2009, when entering the P3R17 outage, This issue was considered as a condition prohibited by TS since there was evidence that the condition had existed during plant operations, The cause of the event was due to not requiring preventive maintenance for the MSIV oil dashpot needle control valve, PBAPS documented this issue in the CAP as IR 964717, Since PBAPS analysis concluded this condition did not have a significant effect on the safety analysis and the plantnever operated outside of the safety analysis, this issue is of very low (Green)safety significance, The LER associated with the event was documented in Section 40A3,2 of this report.
05000277/FIN-2010003-01	2010Q2	Green		Licensee-identified	Licensee-Identified Violation	License Condition 2.C.(11).(b) requires, in part, that PBAPS shall develop and maintain strategies for addressing large fires and explosions. On May 11, 2010, the licensee discovered that equipment used for a single mitigation strategy, described in TSG-4.1, Attachment 15, had been removed from its designated location. The finding was assessed in accordance with IMC 0609, Appendix L, Table 2, and determined to be of very low (Green) safety significance because the finding only impacted an individual mitigation strategy. The licensee restored the equipment and entered the issue into their CAP as IR 01068128.
05000277/FIN-2010004-01	2010Q3			NRC identified	Non-conservative TS and Potential Non-compliance Associated with Degraded SFP Boraflex Panels	The inspectors identified an unresolved item (URI) related to issues of concern with the degrading Boraflex panels in the PBAPS SFPs. Additional information and specialized technical support from the NRC\\\'s Office of Nuclear Reactor Regulation (NRR) are required to determine whether a performance deficiency exists. Specifically, NRR will be requested to provide a technical review of the PBAPS\\\'s operability determination ((00) 10-007) to determine if it is technically sufficient and to confirm the time limitations associated with the referenced technical evaluation. This will support an evaluation of whether PBAPS\\\'s corrective actions to address the non-conservative TS (4.3.1.1.a) associated with the design limit for peak in-core reactivity (k-infinity) of spent fuel have been timely when judged against the standards established in NRC Administrative Letter (AL) 98-10, Dispositioning of TSs That Are Insufficient To Assure Plant Safety, and the requirements in 10 CFR 50, Appendix B, Criterion XIV, Corrective Actions. Additionally, the inspectors will use the results of the NRR technical review to determine whether the PBAPS 00 has demonstrated with reasonable assurance that the subcritical margin limit for the SFP as specified by TS 4.3.1.1.b (K-effective\\\': 0.95) will continue to be met through the time limit established in the technical evaluation and until the licensee\\\'s specified corrective actions can be completed. The current technical evaluation concludes that with administrative limits on the reactivity of the fuel added to SFPs, K effective will conservatively remain below 0.95 until approximately 2014. Since 1996, PBAPS has known that the Boron-10 (B-10) neutron absorber used in the Units 2 and 3 SFPs\\\' racks had begun a degrading trend. Specifically, the degradation caused some of the Boraflex neutron absorber material imbedded in the rack panels to fall below the minimum certified B-10 density of 0.021 grams B-10 per square centimeter (g/cm2). The panels had degraded from the as-manufactured average areal density of 0.0235 g/cm2that was 11.9 percent greater than minimum certified density. In response to degrading trends, PBAPS secured analyses from AEA Technology and NET Co that quantified the reactivity effects associated with varying degrees of B-1 0 density loss in the Westinghouse racks. The reactivity penalty derived from this analysis was transposed into Global Nuclear Fuel (GNF) SFP criticality analyses. PBAPS asserted that these analyses were incorporated into the plants\\\' licensing and design bases through the 10 CFR 50.59 process. However, none of these methods have been reviewed and approved by the NRC for application at Peach Bottom. In 2007, PBAPS recognized that the B-1 0 degradation of the Units 2 and 3 SFPs storage was projected to exceed the 10 percent loss limit (0.0189 g/cm2) established by the AEA Technology, NETCo, and GNF analytical methods. PBAPS also recognized that the K infinity value in TS (4.3.1.1.a) would become non-conservative and the guidance in NRC AL 98-10, Dispositioning of TSs That Are Insufficient to Assure Plant Safety, would apply. Subsequently, PBAPS submitted a license amendment request (LAR) to change the Kinfinity value in the TS. In response to issues raised by the NRC\\\'s technical reviewers, PBAPS made several supplemental submittals to the LAR before it was withdrawn by a letter dated June 18, 2010 (ML 101690377). Subsequently, PBAPS developed 00 10-007 to address the non-conservative TS (4.3.1.1.a). The 00 evaluated the acceptability of storing fuel bundles in the Unit 2 and 3 SFP storage racks with a minimum B-10 average areal density of 0.01155 gm/cm2, which is 55% of 0.021g/cm2 (45% degradation). In comparison, it is noted that the most degraded panel in either units\\\' SFP storage racks was measured in January 2010, to be degraded to an areal density of 0.0169 g/cm2 (19.5 percent of 0.021g/cm2 ) and has been projected to have degraded to 0.0146 g/cm2 (30.5 percent of 0.021g/cm2 ) on November 1, 2010. The degradation projections have been made by the RACKLIFE version 2.0 computer modeling program; however, it is noted that the licensee plans to convert to version 2.1 of RACKLIFE program. The OD referenced and relies on Revision 3 of a technical evaluation (IR 864431-15, and two previous revisions) that PBAPS has used since 2009 to justify continued operability of the SFPs and to show that the SFP will be maintained 5% subcritical (Keff $ 0.95). The basis for the approach in these documents was to reduce the design basis limiting fuel assembly reactivity to a maximum Kinfinily of 1.26. The current technical evaluation concludes that with administrative limits on the reactivity of the fuel added to SFPs, K effective will conservatively remain below 0.95 until the maximum B-10 density depletion reaches approximately 45 percent in 2014. As an additional compensatory measure, PBAPS plans to remove from service any SFP storage rack panels with Boraflex degraded more than 45 percent. PBAPS\\\'s current plans are to submit a new LAR in late 2011. The inspectors reviewed OD 10-007 and concluded that assistance from NRR was needed to determine the technical adequacy and correctness of the licensee\\\'s operability evaluation and to confirm the time limitations associated with the referenced technical evaluation. This assistance is needed by the region to determine whether one or more performance deficiencies exist. Specifically, to evaluate whether PBAPS\\\'s corrective actions to address the non-conservative TS (4.3.1.1.a) associated with the design limit for peak in-core reactivity (k-infinity) of spent fuel have been timely when judged against the standards established in NRC AL 98-10, Dispositioning of TSs That Are Insufficient To Assure Plant Safety, and the requirements in 10 CFR 50, Appendix B, Criterion XIV, Corrective Actions. Additionally, the inspectors will use the results of the NRR technical review to determine whether the PBAPS OD has demonstrated with reasonable assurance that the subcritical margin limit for the SFP as specified by TS 4.3.1.1.b (K-effectives. 0.95) will continue to be met through the time limit established in the technical evaluation and until the licensee\\\'s specified corrective actions can be completed. The inspectors plan to submit their technical questions to NRR in accordance with Office Instruction, COM-106, Control of Task Interface Agreements. Therefore, this issue remains unresolved pending NRR\\\'s response to the TIA and subsequently inspector review. URI 05000277, 278/2010004-01, Non-conservative TS and Potential Noncompliance Associated with Degraded SFP Boraflex Panels.
05000277/FIN-2010004-02	2010Q3	Green		NRC identified	Potentially Inadequate Fuel Handling Procedures Lead to Personnel Performance Errors While Handling Fuel	The inspectors identified an URI related to potential procedure inadequacy issues that allowed inadequate coordination of simultaneous close proximity activities within the reactor vessel and personnel performance error issues while handling fuel in the reactor core and the SFP. These events appear to be examples where inadequate procedures contributed to fuel handling issues. This issue will remain unresolved pending completion of PBAPS\\\'s investigation and cause evaluation processes under the CAP. On September 18, 2010, during Core Shuffle I, the safety spotter had to stop the refueling bridge to avoid contact with the CS inspection (CSI) submarine. On September 19,2010, during the execution of fuel move 302 of Core Shuffle I, a discharged fuel bundle (JLM491), that had been picked up from the core, came in contact with the CSI submarine as the refueling bridge began transiting to the SFP (IR 1115041). Both fuel movement and NOEs using a remotely operated vehicle (CSI submarine) were being conducted within the same core quadrant. On September 24, 2010, during preparations for Core Shuffle II, a dummy fuel bundle came in contact with a discharged fuel bundle at location JJ-37 in the SFP while the refueling bridge\\\'s mast was being lowered over an occupied storage cell using the travel override pushbutton (IR 1115041). At the time, the mast was being exercised in accordance with a refuel bridge ST. At the end of the inspection period, PBAPS\\\'s causal analysis activities were still in progress; therefore, this item remains unresolved: URI 05000277, 278/2010004- 02, Potentially Inadequate Fuel Handling Procedures Lead to Personnel Performance Errors While Handling Fuel.
05000277/FIN-2010004-03	2010Q3	Green		NRC identified	Failure to Ensure Adequate Voltage was Available to Safety-Related Equipment	The inspectors identified a finding of very low safety significance involving a NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, in that Exelon did not assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, Exelon did not use the safety-related Function 4 degraded grid relay trip setpoint specified in the Technical Specifications (TS) as a design input in calculations to ensure adequate voltage was available to all safety-related components required to respond to a design basis loss-of-coolant accident (LOCA). Instead, Exelon used the results from calculation PE 0121, Voltage Regulation Study, to establish the voltage level for system operability. The study credited the use of non-safety related equipment to raise the voltage level. This allowed higher voltages to be used in the design calculations for components than would be allowed by the TS setpoint. The team verified the licensing basis via Task Interface Agreement (TIA) 2009-07 and informed Exelon that the degraded grid relay setpoint must be used for design basis calculations. Exelon entered the issue into the CAP (IR 1119440), performed operability assessments, and established some compensatory measures to restore PBAPS to an operable but nonconforming condition. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was also similar to example 3j in IMC 0612, Appendix E, in that there was reasonable doubt as to the operability of safety-related components and Exelon was required to perform operability determinations to address potentially inadequate voltage to several safety-related components. The inspectors, including the Region I Senior Reactor Analysts (SRAs), performed a Phase 1 SOP screening, in accordance with NRC IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance (Green) because it was a design deficiency that impacted operability but not functionality, did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. There was no cross-cutting issue associated with the finding because the degraded grid relay setpoints had been most recently evaluated in 2004 and the issue was not reflective of current performance.
05000277/FIN-2010009-01	2010Q2	Severity level IV		Self-revealing	Inaccurate Personnel History Questionnaire	a former contract outage employee at Peach Bottom deliberately failed to disclose on a Personal History Questionnaire (PHQ), a previous, non-nuclear employment from which he had been terminated for a positive FFD test, in order to gain unescorted access (UA) to Peach Bottom. As a result of the investigation, the NRC determined that, on September 8, 2008, the contract employee did fail to disclose his prior employment with the non-nuclear company on the PHQ, and also failed to provide information about the positive FFD test. However, after considering the information developed during the investigation, the NRC concluded that it did not have sufficient evidence to conclude that the individuals failures were deliberate. Nonetheless, as a result of these failures by the contract employee, Exelon granted the individual UA to Peach Bottom from September 11, 2008, until September 28, 2008. Exelon learned of the individuals positive FFD in August 2009, when the contract employee attempted to gain UA to Progress Energys Crystal River Nuclear Generating Plant 3 (Crystal River) Although the contract employee did not enter any Vital Areas at Peach Bottom and also did not perform work on any safety-related equipment during the time he was granted access, the contract employees actions caused Exelon to be in violation of NRC requirements, specifically: 1) 10 CFR 50.9, which states in part that information required by the Commissions regulations, orders, or license conditions to be maintained by the licensee shall be complete and accurate in all material respects; and, 2) 10 CFR 73.56(c) and Section 9.1 of the Peach Bottom Physical Security Plan, both of which state, in part, that the licensees access authorization program must provide high assurance that the individuals who are granted unescorted access are trustworthy and reliable. Although Exelon was unaware of the contract employees omission of information regarding the positive FFD test, Exelon is responsible for the adequacy of its Physical Security Plan and background checks to identify past actions and appropriately evaluate the trustworthiness and reliability of applicants for UA. (This item was also discussed in Inspection Report 2010-004.)
05000277/FIN-2011002-01	2011Q1	Green	H.11 H.12	Self-revealing	FH Procedures Were Inadequate to Prevent Fuel from Contacting an Obstruction	A Green self-revealing NCV of Technical Specification (TS) 5.4.1 Procedures was identified, because PBAPS\'s procedures for refueling equipment operation and core alterations were inadequate to prevent a fuel bundle from contacting a core spray inspection (CSl) submarine device while the fuel bundle was being transported from the core to the spent fuel pool (SPF). In particular, system operating (SO) procedure 18.1.A-2, Operation of Refueling Platform, and fuel handling (FH) procedure 6C, Core Component - Core Transfers, did not provide sufficient procedure steps, precautions, or human performance tools to prevent contact while the refueling platform was operated in the automatic mode and when core components were in close proximity to obstructions and interferences. The inspectors determined that the finding was more than minor because the finding was associated with the Procedure Quality attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone\'s objective to provide reasonable assurance that physical design barriers (i.e., fuel cladding) protect the public from radionuclide releases caused by accidents or events. Although no fuel damage occurred during this event, the inadequate procedure resulted in a FH event that could have impacted the cladding and affected the cornerstone\'s objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. IMC 0609, SDP, Attachment 0609.04, Phase 1-lnitial Screening and Characterization of Findings, was used to evaluate the significance of the finding. Attachment 0609.04, Table 4a, was used to evaluate the impact of the finding on fuel clad integrity. Appendix G was considered for the evaluation, but was not used because it does not directly address fuel clad integrity. Based on the results of fuel sipping done in February 2011, PBAPS concluded that there was no damage to the clad integrity of the impacted fuel bundle that was permanently discharged to the SFP. Since the finding did not affect SFP cooling or inventory and since there was no damage to fuel clad integrity from the impact with the CSI submarine, the finding was determined to be of very low safety significance (Green). The finding has a cross-cutting aspect in Human Error Prevention Techniques in the Work Practices component of the Human Performance area. Specifically, PBAPS FH procedures did not require human error prevention techniques that were commensurate with the risk of moving fuel in close proximity to obstructions and interferences. (H.4(a))
05000277/FIN-2011002-02	2011Q1	Green		Licensee-identified	None	In Modes 1, 2 and 3, with one ESW subsystem inoperable for more than seven days, TS Limiting Condition for Operation (LCO) 3.7.2, condition C, requires the unit to be in Mode 3 within 12 hours and in Mode 4 within 36 hours. Contrary to the above, since original construction and prior to September 13, 2010, an engineering evaluation determined that the \'A\' ESW subsystem was inoperable due to the degraded seismic capability of rod hanger 33HB-S143 that only affected the \'A\' ESW subsystem. During upgrades to the ESW discharge pipe support system during the week of September 13, 2010, PBAPS personnel identified that the original installation of the rod hanger had not been carrying adequate pipe load. This condition was considered as a condition prohibited by TS due to one subsystem of ESW being inoperable for greater than the time period allowed by TS. The cause of the event was due to an inadequate design drawing. PBAPS documented this issue in the CAP as IRs 1114812 and 1118711. Since there was no actual Joss of safety function as a result of this event, this issue is of very low (Green) safety significance. The LER associated with the event was documented in Section 40A3.1.
05000277/FIN-2011002-03	2011Q1	Green		Licensee-identified	None	TS LCO 3.4.3 requires the safety function of 11 valves (any combination of SRVs and SVs) to be operable during operational Modes 1,2, and 3 or else be in Mode 3 within 12 hours and in Mode 4 within an additional 36 hours. Contrary to the above, two SRVs and one SV were determined to have their as-found setpoints in excess of the TS allowable tolerance, thus leaving 10 operable SRVs and SVs. The SRVs and SVs were replaced with refurbished valves for the 19th Unit 2 operating cycle. Additionally, LER 2-10-3 stated that PBAPS will pursue a change to the plant\'s licensing bases to increase SRV and SV setpoint tolerances to the ASME Code allowable + 3 percent tolerance. The licensee documented the event in JR 1120516. Since there was no actual loss of safety function as a result of this event, this issue is of very low (Green) safety significance. The LER associated with the event was documented in Section 40A3.2.
05000277/FIN-2011003-01	2011Q2	Green		Licensee-identified	Licensee-Identified Violation	In Mode 1, with the HPCI system inoperable for more than 14 days, TS Limiting Condition for Operation 3.5.1 requires the unit to be in Mode 3 within 12 hours. Contrary to the above, the Unit 2 HPCI system was determined to be inoperable from approximately January 20 to March 18, 2011, with the reactor in Mode 1, due to a leaking relief valve (RV-2-238-066) on the HPCI cooling water header. With HPCI aligned to the normal, non-safety-related, Condensate Storage Tank (CST) suction source, no voiding would occur in the HPCI discharge piping due to the higher elevation of the CST. However, during a subset of design basis events where HPCI suction would be transferred to the suppression pool, its alternate and safety-related suction source, and the HPCI pump secured, voiding could develop in the discharge piping. The licensee concluded that if HPCI was then restarted, a water hammer condition could potentially result and render Unit 2 HPCI unable to perform its deterministic design function. The voiding in the HPCI discharge piping had been discovered by PBAPS personnel during a ST while transferring Unit 2 HPCI suction from the CST to the suppression pool to support an l&C surveillance. The relief valve was replaced, and subsequent to testing, HPCI was declared operable on March 18, 2011. The inspectors reviewed this condition using IMC 0609, Attachment 4, and in consultation with a Region I Senior Reactor Analyst (SRA), concluded the Unit 2 HPCI system would likely have been able to perform its Significance Determination Process safety function, given the numerous postulated equipment failures and specific system configurations that would have to occur to cause a system failure. Therefore, and as such this issue screened to very low safety significance. A Region I SRA also confirmed the very low significance (mid E-9 increase in core damage frequency) with a conservative analysis. This analysis assumed the HPCI system would have failed if the operators failed to refill the CST, and HPCI switched over to the torus suction, for the 58 day exposure period. The licensee documented the event in their CAP as lRs 1 1 88457 and 1 188987. The LER associated with this event was documented in Section 4OA3.
05000277/FIN-2011004-01	2011Q3	Green		Licensee-identified	Licensee-Identified Violation	10 CFR, Part 50, Appendix B, Criterion III requires, in part, that measures shall be established to assure that the design basis for those SSCs that mitigate the consequences of postulated accidents are correctly translated into procedures. Contrary to the above, PBAPS did not ensure that the CS system required flow of 6,874 gallons per minute (gpm) was correctly translated into Emergency Operating Procedure T-111, Level Restoration, to ensure long term core cooling following a loss of coolant accident. The 6,874 gpm flowrate was determined by engineering analysis to account for the 624 gpm leakage through the CS sparger headers and into the reactor vessel annulus region, thereby bypassing long-term cooling of the fuel in the core shroud region. The inspectors determined that this finding was of very low safety significance (Green) in ac ordance with NRC IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, Mitigating Systems cornerstone, because the finding did not result in the actual loss of safety function. PBAPS engineering review of quarterly surveillance tests for the last three years determined that the CS pumps have more than sufficient margin to account for the leakage. The inspectors verified the determination through an independent inspection sampling of surveillance test data. This finding has been documented in I the CAP under IR 1245207.
05000277/FIN-2011005-01	2011Q4	Green	P.3	NRC identified	Untimely Corrective Action to Correct MOV Degraded Stem Lubrication	The inspectors determined that Exelon\'s failure to promptly correct a condition adverse to quality associated with a safety-related motor-operated valve (MOV) constituted a Green, self-revealing NCV of 10 CFR Part 50, Appendix B, Criterion XVl, Corrective Action. Specifically, corrective actions to prevent recurrence of MOV program testing failures due to degraded stem lubrication in 2009 were not performed in a timely manner to prevent the inoperability of a safety-related MOV due to degraded lubrication, as identified on September 22, 2011. PBAPS entered this issue into the CAP via issue reports (lRs) 1266600 and 1266604. This finding was more than minor because it was associated with the configuration control attribute of the Barrier Integrity (Bl) cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the Unit 3 reactor water cleanup (RWCU) outboard isolation valve MO-3-12-018 did not develop sufficient thrust at the torque switch trip setpoint during diagnostic testing on September 22, 2011. The RWCU MOV would not have been able to perform its safety function to close during the most limiting design condition. Using the Phase \'1 worksheet in Appendix 4 of IMC 0609, SDP, the finding affected the Bl cornerstone and was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of containment. This finding had a cross-cutting aspect in the area of Problem ldentification & Resolution (PI&R), CAP, because Exelon did not take appropriate corrective actions to address the adverse trend of degraded stem lubrication on a safety-related MOV in a timely manner
05000277/FIN-2011005-02	2011Q4	Green	H.8	NRC identified	Failure to Establish, Implement, and Maintain Adequate QA for Effluent and Environmental Monitoring	The inspectors identified a Green finding associated with the failure to establish, implement, and maintain adequate quality assurance (QA) program elements in the area of effluent and environmental monitoring as required by Peach Bottom, Units 2 and 3 Technical Specification (TS), Section 5.4.1. Specifically, Exelon\\\'s QA program for effluent and environmental monitoring was not sufficient to ensure: 1) that both adequate and timely evaluation and assessment of changes described in the Public Land Use Census were conducted for purposes of dose validation and sampling program modification; 2) that changes in meteorological parameters, used for public dose projections and assessment, were promptly and adequately evaluated; and 3) that laboratory QA programs for effluent and environmental sample analysis measurement systems were adequate and implemented properly. Exelon placed these issues in its CAP as Action Requests (ARs): 1226969, 1226202,1299543, 1299476,1302720, and 1303308. The finding is more than minor because it is associated with the Public Radiation Safety cornerstone attribute of programs and processes and adversely affected the associated cornerstone objective in that failure to establish, implement, and maintain an adequate QA program in the effluents and environmental monitoring program area adversely affected the licensee\\\'s ability to ensure adequate protection of public health and safety. The finding was assessed for significance using IMC 0609, Appendix D, and determined to be of very tow safety significance (Green) because: the issue was contrary to TSs and is a radioactive effluent release program deficiency; there was no indication of a spill or release of radioactive material on the licensee\\\'s site or to the offsite environs that would impact public dose assessment, and there was no substantial failure to implement the radioactive effluent release program. The licensee re-assessed the dose to members of the public from routine releases and determined that projected doses did not, nor were likely to, exceed applicable limits, including as low as is reasonably achievable (ALARA) design specifications of 10 CFR Part 50, Appendix l; or 10 CFR 20.1301(e). The cause of this finding is related to the cross-cutting area of Human Performance, Work Practices, Aspect H.4(b) because the licensee did not ensure Personnel followed procedure compliance requirements activities for effluent and environmental monitoring program.
05000277/FIN-2011005-03	2011Q4	Green		Licensee-identified	Licensee-Identified Violation	TS b.4.1 states, in part, that written procedures shall be implemented a1d maintained as recommended in RC 1.33, Appendix A, November 1972. RG 1\'33, Appendix A\' Section l, procedures for Performing Maintenance, subsection 1, states the following: Maintenance which can affect the performance of safety-related Equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances\' Skill, notably possessed by qualified maintenance personnel may not require detailed step--by-step delineation in a procedure. Contrary to the above,. PBAPS did not properly preplan and perform maintenance which affected the E-1 EDG\' Specifically, PBAPS determined that a damaged lubricating oil drain line should have been identified and replaced during planned maintenance activities prior to the occurrence of leakage. As a consequence of not identifying and replacing the damaged drain line, PBAPS determined that the E-1 EDG was unable to perform its 24-hour mission time, and therefore was inoperable, during the period of time between April27 , 2011, and September 23, 2011\' The finding was determined to be of very low safety significance, for both Peach Bottom Units 2 and 3, in accordance with lMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations (lMC 06094) using SDP phase s 1,2 and 3. Phase 1 screened the finding to Phase 2 because it represented a loss of the E-1 EDG safety function, between April27 and September 23, 2011 (149 days), longer than the TS limiting condition for operation (LCO) of 14 days. A Region I senior Reactor Analyst (SRA) conducted a Phase 3 analysis because the Phase 2 analysis, conducted by the inspectors using- the Peach Bottom pre-solved Risk-Informed Inspection Notebook, indicated that the finding could be more than very low significance\' The SRA used the peach Bottom Standardized Plant Analysis Risk (SPAR) model, Revision 8.19 and 8.17, for units 2 and 3 respectively and SAPHIRE 8 to conduct the phase 3 analysis, with the conservative assumption that the E-1 EDG would not have operated at all for its 24 hour mission time over the 149 day exposure period\' This analysis was conservative given the EDG could have operated for over two hours assuming that the drain line broke and the potential that operators could have temporarily limited the leakage from the supercharge lube oil drain line. This analysis indicated an increase in core damage frequency (CDF) for internal initiating events in the range of one core damage accident in 2,500,000 years of reactor operation, in the low E-7 range per year for each unit. The dominate core damage sequences included losses of offsite power with the failure of all EDGs resulting in a station blackout (SBO), followed by the failure of operators to reduce direct current loading to allow extended operation of the RCIC system and - depressurize the realtor, and with inability to recover offsite power, the SBO source of power from the Conowingo Dam or an EDG in two hours. In accordance with IMC 0609A, for a finding with an internal events ACDF above 1E-7, the SRA assessed the impact of the finding on: 1) External events such as fire, seismic and flooding, determining, using the external events portion of the Peach Bottom Unit 2 and 3 SPAR models, that the total ACDF (internal plus external) would not be above the 1 E-6 threshold; and 2) the increase in large early release frequency (ALERF)\' determining that given the operators ability, following core damage, to recover offsite power and depressurize and inject water to the reactor from low pressure sources and to flood the containment that the ALERF was in the low E-8 per year range\' Because this finding is of very low safety significance and has been entered into Exelon\'s CAP under lR 1266b37, this violation is being treated as a Green, licensee identified NCV consistent with the NRC Enforcement Policy.
05000277/FIN-2011005-04	2011Q4	Green		Licensee-identified	Licensee-Identified Violation	TS LCO 3.5.1, Condition A, requires that one inoperable low pressure ECCS injection subsystem should be restored to an OPERABLE status within seven days during operational modes 1 and 2, or requires action to place the unit in operational mode 3 within 12 hours. Contrary to the above, the \'D\' LPCI pump was inoperable during a period of time between April27, 2O1O, and October2, 2011. Specifically\' PBAPS determined that the leaking relief valve body, as identified on April 27,2010, could have become detached from the \'D\' RHR suction piping during the worst case design basis seismic event. This condition would result in the \'D\' RHR pump being inop6rable, thereby affecting the RHR LPCI function. Because the \'B\' RHR pump was unaffected by this even-t, there was no total loss of the \'B\' LPCI train safety function. The inspectors determined that this event screens to Green using the Table 4b seismic screening criteria in Attachment 4 of IMC 0609, SDP. Because this finding is of very low safety significance and has been entered into Exelon\'s CAP under lR i264g09, this violation is being treated as a Green, licensee-identified NCV consistent with the NRC Enforcement Policy
05000277/FIN-2011007-01	2011Q1	Green		NRC identified	Failure to Demonstrate the Capability of the EDG Fuel Oil Transfer Pumps to Fulfill Their Safety Functions Under all Postulated Conditions	The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Exelon did not ensure the ability to transfer fuel oil between underground fuel oil storage tanks. Specifically, Exelon had not performed adequate analyses or testing to demonstrate adequate net positive suction head available (NPSHA) for the EDG fuel oil transfer pumps. In response, Exelon entered this issue into their corrective action program and performed an evaluation to assure the fuel oil transfer pump NPSHA was adequate. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team performed a Phase 1 SOP screening, in accordance with NRC IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. This finding does not have a crosscutting aspect because the most significant contributor of the performance deficiency is not reflective of current licensee performance.

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