Text

IR 05000219/2014010, December 8, 2014 Through 11, 2014, Oyster Creek, NRC Supplemental Inspection Report
	ML15020A632
Person / Time
Site:	Oyster Creek
Issue date:	01/20/2015
From:	Kennedy S R; NRC/RGN-I/DRP/PB6
To:	Bryan Hanson; Exelon Generation Co, Exelon Nuclear
	KENNEDY, SR
References
	IR-2014010
	Download: ML15020A632 (19)
	v • d • e

January 20, 2015

Mr. Bryan Hanson Senior Vice President, Exelon Generation Company, LLC President and Chief Nuclear Officer, Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT: OYSTER CREEK NUCLEAR GENERATING STATION - NRC SUPPLEMENTAL INSPECTION REPORT 05000219/2014010

Dear Mr. Hanson:

On December 11, 2014, the U. S. Nuclear Regulatory Commission (NRC) completed a supplemental inspection pursuant to Inspection Procedure 95001, "Supplemental Inspection for One or Two White Inputs in a Strategic Performance Area," at your Oyster Creek Nuclear Generating Station. The enclosed inspection report documents the inspection results, which were discussed on December 11, 2014, with Mr. J. Dostal, Plant Manager, and other members of your staff. performance indicator (PI) crossed the green-to-white threshold following an unplanned scram -to-white. Based on your report, the NRC assigned a white PI Action Matrix input to the Initiating Events cornerstone in the third quarter of 2014, due to unplanned scrams on October 3, October 6, and December 14, 2013, and July 11, 2014. In response to this Action Matrix input, the NRC informed you in our mid-cycle assessment letter dated September 2, 2014, that a supplemental inspection under Inspection Procedure 95001 this inspection. The NRC performed this supplemental inspection to determine if (1) the root causes and the contributing causes for the risk-significant issues were understood; (2) the extent of condition and extent of cause of the issues were identified; and (3) corrective actions were or will be sufficient to address and preclude repetition of the root and contributing causes. The inspection consisted of examination of activities conducted under your license as they related to safety, s, and the conditions of your operating license. Based on the results of this inspection, the NRC concluded that, overall, the supplemental inspection objectives were met and no significant weaknesses were identified. The NRC evaluation identified a collective root cause of the four reactor scrams which was decision makers do not always understand the likelihood or consequence of the malfunction of degraded equipment to cause a transient. To correct this root cause, Exelon developed a single risk process to help decision makers better understand the risk and consequences of malfunctions which cause plant transients. The NRC has determined that completed or planned corrective actions were sufficient to address the causes of the events that led to the white PI. In accordance with Inspection Manual Critical Hours performance indicator will continue to be considered as a White Action Matrix input until the performance indicator has returned to the Green performance band. Any future changes in Action Matrix column designation will be communicated via separate correspondence. One self-revealing finding of very low safety significance (Green) was identified. This finding did not involve a violation of NRC requirements. If you disagree with the cross-cutting aspect or the finding not associated with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I; and the NRC Resident Inspector at Oyster Creek. a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records t System (ADAMS). ADAMS is accessible from the NRC website at http://www.nrc.gov/reading rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/ Silas R. Kennedy, Chief Reactor Projects Branch 6 Division of Reactor Projects Docket Nos. 50-219 License Nos. DPR-16

Enclosure:

Inspection Report 05000219/2014010

w/Attachment:

Supplementary Information cc w/encl: Distribution via ListServ

SUMMARY

IR 05000219/2014010; 12/8/2014 12/11/2014; Oyster Creek Nuclear Generating Station; Supplemental Inspection Inspection Procedure (IP) 95001 A regional Senior Project Engineer and a Resident Inspector from the Division of Reactor Projects, Region I, performed this inspection. No significant weaknesses were identified in this report. One self-revealing finding of very low safety significance (Green) was identified. The significance of most findings is indicated by their color (i.e., greater than Green, or Green, ated June 2, 2011. Cross-cutting aspects are determined -described in NUREG-1649, Inspection Procedure 95001, "Supplemental Inspection for One or Two White Inputs in a The NRC performed this supplemental inspection in accordance with NRC inspection procedure 95001, n-to-white threshold (value > 3.0) on July 11, 2014, when Oyster Creek experienced its fourth reactor scram in the previous ten months of operation. Based on the results of the inspection, the inspectors concluded that Exelon had adequately performed a root cause evaluation (RCE) or apparent cause evaluation (ACE) for each event, and a RCE collectively for the four events. The NRC has determined that completed or planned corrective actions were sufficient to address the causes of the events that led to the white PI.

Cornerstone: Initiating Events

Green.

A self-revealing finding (FIN) of very low safety significance was identified for temporary repair was performed on condenser bellows expansion joint Y-1-26. The temporary repair impacted the design function of Y-1-26 and led to failure of the downstream side of the bellows, causing a loss of condenser vacuum and manual reactor scram on July 11, 2014. Exelon replaced both the expansion joint Y-1-26 and the 2nd stage reheater steam supply relief valve V-1-132 on July 11, 2014, during forced outage 1F35. Exelon entered this issue into the corrective action program (IR 2422831). This finding was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined that this finding was of very low safety significance (Green) using Exhibit 1 of Findings At-ause the finding did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser, loss of feed water). The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Exelon did not systematically and effectively evaluate relevant internal operating experience related to a similar condenser bellows expansion joint failure in 1986. [P.5] (Section 4OA3)

REPORT DETAILS

4.

OTHER ACTIVITIES

4OA3 Follow-Up of Events and Notices of Enforcement Discretion

Main article: IP 71153

.1 (Closed) Licensee Event Report (LER) 05000219/2013-004-00:

Manual Scram due to Rise in Reactor Pressure during Turbine Valve Testing On December 14, 2013, operators initiated a manual reactor scram due to reactor pressure control abnormalities during quarterly turbine valve testing with reactor power at 95 percent of rated thermal power. Turbine control valves 2 and 3 failed closed when the servo motor feedback support bracket bolts loosened, then detached, from their supports. Also, a vertical connection to transmit the required turbine bypass valve relay position from the turbine front standard to the bypass valve assembly had become detached. This condition is reportable under Title 10 of the Code of Federal Regulations (10 CFR) 50.73(a)(2)(iv)(A) as an event that resulted in a manual actuation of the reactor protection system.

A RCE was performed to evaluate the equipment failures and determined the root cause was the original equipment manufacturer during manufacture did not follow their assembly drawings and installed inappropriate locking mechanisms (split washers) instead of the assembly drawing required parts (lock plates). The corrective action restored the turbine control valve servo motor feedback support brackets to original assembly design by installing lock plate securing mechanisms.

The inspectors performed an in-documentation, station procedures, and interviewed members of station staff and management regarding the event. No issues were identified during this review. This LER is closed.

.2 (Closed) LER 05000219/2014-001-00:

Manual Scram due to Lowering Vacuum

a. Inspection Scope

On July 11, 2014, during planned reactor power ascension with reactor power at approximately 56 percent of rated thermal power, operators initiated a manual reactor scram upon receiving indications of degrading main condenser vacuum. The apparent cause of the event was determined to be failure of the downstream bellows of the 'B' Condenser Steam Inlet Expansion Joint due to fatigue loading. On October 6, 2013, the upstream bellows was repaired due to a circumferential fracture. The upstream fracture in the bellows was repaired with standard fiberglass wraps, high temperature carbon fiber wraps, and the application of Belzona. Repeated wrapping of the upstream side of the bellows most likely restricted the allowable movements of that bellows, requiring the downstream bellows to account for the additional movement. Additionally, in July 2014, a reheater relief valve (V-1-132) upstream of the bellows was confirmed to be leaking past its seat. The combination of the leak-by of the relief valve with the restricted movement of the bellows created increased fatigue stresses on the downstream bellows.

This condition is reportable under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in apparent cause analysis, supporting documentation, station procedures, and interviewed members of station staff and management regarding the event. A self-revealing finding of very low safety significance (Green) was identified and is discussed below. This LER is closed.

b. Findings

Introduction.

A self-revealing finding of very low safety significance (Green) was identified for Exewhen a temporary repair was performed on condenser bellows expansion joint Y-1-26. The temporary repair impacted the design function of Y-1-26 and led to failure of the downstream side of the bellows, causing a loss of condenser vacuum and manual reactor scram on July 11, 2014.

Description.

On October 6, 2013, Oyster Creek operators entered abnormal operating procedure (ABN)-14 due to degrading condenser vacuum. A manual scram was later upstream side of Y-1-function of Y-1-26 is to provide a flexible pressure retaining connection to absorb motion in a system caused by thermal expansion and low levels of vibration. Y-1-26 is a universal type expansion joint which contains two bellows in series separated by a pipe spool and tie rods designed to contain the pressure thrust force.

likely due to fatigue cracking. An indentation at the fracture site also suggested an impact had occurred, possibly during a past maintenance outage. A condenser waterbox preventive maintenance inspection task was completed in October 2012 during refueling outage 1R24, during which an accidental impact from a tool or scaffold pole may have occurred. Rather than replace the bellows, Exelon decided on October 7, 2013, to perform a temporary repair by wrapping the bellows circumferentially with fiberglass and then applying Belzona, a type of industrial coating. After the temporary repair was completed, the plant was restarted. Between October 7, 2013, and July 9, 2014, the bellows repair was rewrapped a total of five times. The repeated wrapping of the bellows stiffened and restricted the normal design movement of the bellows, and as a result, transferred higher fatigue loading and stresses to the downstream bellows.

On November 17, 2013, operators identified leak-by of 2nd stage reheater steam supply relief valve V-1-132. Steam from V-1-passes through the Y-1-26 condenser bellows expansion joint. Recognizing the challenge to the integrity of the temporary leak repair caused by this leak-by, system engineers scheduled the replacement of V-1-132 in the next forced outage. On July 11, 2014, the combination of the V-1-132 leak-by and higher stresses caused by the temporary leak repair wrapping led to a failure of the downstream bellows, a loss of condenser vacuum, and subsequent manual reactor scram. Exelon replaced both the expansion joint Y-1-26 and the 2nd stage reheater steam supply relief valve V-1-132 on July 11, 2014, during forced outage 1F35. Exelon entered this issue into the corrective action program (IR 2422831). Exelon procedure CC-AA-configuration change as a change to the form, fit, or function of any structure, system or component (SSC). If a temporary installation or alteration of an SSC is performed to allow for continued operation, then a temporary configuration change package is required. By wrapping the bellows with fiberglass and Belzona, Exelon performed a modification to the expansion joint. Specifically, the function of the expansion joint was modified when the temporary leak repair was applied, which restricted the normal design movement of the expansion joint and contributed to its failure on July 11, 2014. In December 1986, Oyster Creek experienced a similar event (LER 86-034-01) when a manual scram occurred due to the lifting of V-1-132, which resulted in a failure of the Y-1-26 bellows expansion joint. Operating experience from this event should have prompted the station to consider taking immediate corrective actions on V-1-132 and/or Y-1-26 before a bellows failure, but this operating experience was not fully considered in Analysisrepair in accordance with station procedure CC-AA-correct, and should have been prevented. This finding was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the higher stresses caused by the temporary leak repair caused a failure of the bellows and led to a manual reactor scram. The inspectors determined that this finding was of very low safety significance (Green) using Exhibit 1 of NRC IMC -the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser, loss of feed water). The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Exelon did not systematically and effectively evaluate relevant internal operating experience related to a similar condenser bellows expansion joint failure in 1986. [P.5]

Enforcement.

This finding does not involve enforcement action because no violation of a regulatory requirement was identified. Because this finding does not involve a violation and is of very low safety significance (Green), it is identified as a FIN. (FIN 05000219/2014010-01, Failure to Evaluate a Temporary Configuration Change)

4OA4 Supplemental Inspection (IP 95001)

.1 Inspection Scope

The NRC conducted this supplemental inspection in accordance with IP 95001, evaluations associated with a white Initiating Events cornerstone PI reported in the third quarter of 2014. The objectives of this supplemental inspection were to:

Provide assurance that the root and contributing causes of risk-significant issues were understood; Provide assurance that the extent of condition and extent of cause of risk-significant issues were identified; and Provide assurance that corrective actions for risk-significant issues were sufficient to address the root and contributing causes and to preclude repetition. The following four reactor scrams contributed to the white PI: On October 3, 2013, at 0643 hours0.00744 days <br />0.179 hours <br />0.00106 weeks <br />2.446615e-4 months <br />, an automatic reactor scram occurred during a plant startup from maintenance outage 1M30. The scram was caused by both Reactor Protection Systems channels receiving simultaneous intermediate range monitor (IRM) Hi-Hi signals on IRM channels 12, 13, 14, 17, and 18. All IRMs began behaving erratically at the same time, cycling between Hi-Hi and downscale, due to noise susceptibility of the IRM channels (Issue Report (IR) 1567196).

On October 6, 2013, during reactor startup from maintenance outage 1M30, main condenser vacuum began to significantly degrade and resulted in a manual reactor scram. The source of the condenser vacuum degradation was a hole of joint (Y-1-bellows led to fracture by fatigue cracking or stress corrosion cracking (IR 1568503).

On December 14, 2013, Oyster Creek initiated a manual reactor scram due to an uncontrolled reactor pressure rise. Troubleshooting identified that turbine control valve 2 and 3 servo motor feedback lever brackets had become loose, then detached, from their supports. A vertical connection to transmit the required turbine bypass valve position from the turbine front standard to the bypass valve assembly had also detached. The original equipment manufacturer did not follow their assembly drawings during manufacturing and installed inappropriate locking mechanisms (split washers) instead of the assembly drawing required parts (lock plates) (IR 1597041).

On July 11, 2014, during reactor start up from forced outage 1F34, Abnormal Procedure (ABN) ABN-14, Loss of Condenser Vacuum, was entered due to rapidly degrading condenser vacuum. Reactor power was lowered in an attempt to stabilize plant conditions. ABN-1, Reactor Scram, was subsequently entered and a manual reactor scram was performed when condenser vacuum degraded to 23.5" Hg. The Y-1-26 expansion joint downstream bellows had failed due to fatigue cracking caused by high stress on the downstream bellows from rigidity of the repeated temporary leak repairs on the upstream bellows and steam leak-by from V-1-132 (IR 1680755).

The Unplanned Scrams per 7000 Critical Hours PI is based on the number of unplanned scrams that are experienced by a unit within the previous 7000 critical hours of reactor operation as measured on a 12-month periodicity. During a time-frame spanning approximately ten months, from October to July 2014, Oyster Creek experienced four reactor scrams. This resulted in plant performance crossing the green-to-white PI threshold value of greater than three unplanned scrams per 7000 critical hours. As a result, the third quarter of 2014. Exelon informed the NRC staff on November 5, 2014, that they were ready for the supplemental inspection. The insscrams (IR 1687264), the causal evaluations conducted for each reactor scram (IRs listed above), and a focused self-assessment completed by Exelon as a readiness review for the NRC supplemental inspection. The inspectors reviewed corrective actions that were taken and planned to address the identified causes. The inspectors also held discussions with Exelon personnel to ensure that the root and contributing causes and the contribution of safety culture components were understood and corrective actions taken or planned were appropriate to address the causes and preclude repetition.

.2 Evaluation of the Inspection Requirements

2.01 Problem Identification a. As directed by IP 95001, determine that the evaluation documented who identified the issue (i.e., licensee-identified, self-revealing, or NRC-identified) and under what conditions the issue was identified. Between October 3, 2013 and July 11, 2014, the Oyster Creek reactor scrammed four times. This resulted in Oyster Creek crossing the green-to-white PI threshold value of greater than three unplanned scrams per 7000 critical hours during the third quarter of 2014. This was properly reported by Exelon to the NRC during the third quarter 2014 Nreporting process, the white PI is considered licensee-identified.

Overall, the documents the identification of the issue and the conditions under which the issue was identified. A timely RCE or ACE was conducted for each individual reactor scram event. Additionally, a collective RCE for the four scrams to identify common causes among the four events was performed. b. As directed by IP 95001, determine that the evaluation documented how long the issue existed and prior opportunities for identification.

The Oyster -to-white threshold value following the July 11, 2014, manual scram, and was properly reported to the NRC via the third quarter 2014 PI submittal. The inspectors determined that Exel(since the first scram in October 3, 2013) and prior opportunities for identification.

c. specific risk consequences, as applicable, and compliance concerns associated with the issue(s). In their collective root cause report (IR 1680755), the inspectors noted that Exelon assessed the risk consequences from four scrams over one year and concluded that the increase in core damage frequency to be approximately 1.3E-6. The December 14, 2013 scram was determined to be complicated due to the failure of the bypass system and required operators to maintain reactor pressure control using the isolation condensers in accordance with emergency operating procedures. Since the PI program performance to be considered in the Reactor Oversight Process vice a regulatory requirement, there are no compliance concerns for the white PI. For the individual scrams, performance deficiencies were identified and a finding of very low safety significance (Green) was documented for the October 3, 2013 reactor scram in NRC Inspection Report 05000219/2014002 and for the July 11, 2014 reactor scram in Section

4OA3 of this report.

As documented in the respective inspection reports, corrective actions were planned or completed to restore compliance. The team reviewed these corrective actions as discussed in Section 2.03. documented the plant specific risk consequences, as applicable, and compliance concerns associated with the issue.

d. Findings

No findings were identified. 2.02 Root Cause, Extent of Condition, and Extent of Cause Evaluation a. As directed by IP 95001, determine that the licensee evaluated the issue using a systematic methodology to identify the root and contributing causes. The inspectors verified that Exelon staff implemented PI-AA-125-contributing causes. The station utilized a variety of causal analysis methods listed in PI-AA-125-included Event and Causal Factor Charts, Failure Analysis, Taproot Analysis and Barrier

Analysis.

The inspectors noted these techniques were supported by data gathering via interviews and document reviews.

The root and apparent causes for the five cause evaluations performed by Exelon are summarized below.

IRM Erratic Behavior Causes Automatic Reactor Scram (IR 1567196) The root cause of the event was determined to be susceptibility of the IRM channels to electrical noise due to low shield to ground resistance. Contributing to the event was an internal fault on the 22 SRM which caused significant noise coupling to occur on the SRM. Exelon determined moisture intrusion into the cabling contributed to the degradation of the IRM ground resistance.

Condenser Expansion Bellows (Y-1-26) Hole and Fracture (IR 1568503) The apparent cause of the bellows failure was impact damage due to inadequate protection of the bellows led to fracture by fatigue cracking or stress corrosion cracking. Contributing to the problem is that the expansion bellows inspection preventive maintenance does not provide guidance for visual inspection of surface defects and corrosion.

Turbine Control System to Control Reactor Pressure (IR 1597041) The root cause determined the original equipment manufacturer (circa 1965) did not follow their assembly drawings during manufacture and installed inappropriate locking mechanisms (split washers) instead of the assembly drawing required parts (lock plates).

Condenser Expansion Bellows (Y-1-26) Failure (IR1680755) The apparent cause of the event was determined to be failure of the downstream side of condenser bellows Y-1-26 due to fatigue loading. A previous temporary leak repair on the upstream side caused the expansion joint to be restricted and unable to account for loading during plant operation. This failure was accelerated by ongoing leak-by of the steam supply relief valve for the 2nd stage reheater V-1-132.

White NRC PI for Unplanned Scrams per 7000 Critical Hours (IR 1687264) The root cause was determined to be decision makers do not always understand the likelihood or consequence of the malfunction of degraded equipment to cause a transient. Contributing to the problem is that internal operating experience is not utilized to determine the likelihood and/or consequence of the malfunction of degraded equipment to cause a transient. b. As directed by IP 95001, determine that thconducted to a level of detail commensurate with the significance of the issue. appropriately implemented their procedures and processes to determine the appropriate causal factors in each of the four reactor scram events. Overall, the inspectors determined tconducted to a level of detail commensurate with the significance of the issues. c. consideration of prior occurrences of the issue and knowledge of operating experience (OE). reviewed OE from multiple sources including the Exelon fleet corrective action program and the Oyster Creek corrective action process. Additionally, relevant NRC generic root and apparent cause teams identified several internal and external OE items that were relevant to the stacause and apparent cause process and corrective actions.

considered relevant OE to inform their investigations and causal determination process. However, in one instance, the inspectors identified that Exelon did not incorporate lessons learned from a 1986 Oyster Creek event when condenser bellows Y-1-26 had failed after V-1-132 began to leak by. Recommendations from the 1986 event to protect the bellows from the introduction or intrusion of seawater during maintenance activities were never incorporated into the site preventive maintenance program. The failure to incorporate the OE was determined not to be a violation of NRC requirements. Exelon entered this issue into their corrective action program (IR 2422831). d. the extent of condition and extent of cause of the issue. Exelon completed individual cause evaluations for each of the four reactor scrams. Additionally, Exelon performed a common root cause evaluation that considered the collective impact of the four reactor scrams that occurred from October 2013 to November 2014.

The team concluded that adequate extent of cause and extent of condition reviews were conducted for each individual reactor scram event as part of their root cause and apparent cause evaluations. Additionally, Exelon also conducted a programmatic and organizational factors review to identify latent organization weaknesses in each of the and apparent cause evaluations addressed the extent of condition and extent of cause of the issue. e. and extent of cause evaluations appropriately considered the safety culture components as described in IMC 0305. Exelon conducted the safety culture reviews in accordance with PI-AA-125-1006, required to have a safety culture review.

Exelon safety culture reviews evaluated the 13 safety culture components in NRC Regulatory Issues Summary 2006-Within the Cross-ppropriately identified station performance gaps with respect to aspects of human performance, decision-making, and corrective action program prior opportunities for identification during its review. Exelon developed corrective actions commensurate to the identified performance gaps to prevent recurrence. consideration of whether the root cause, extent of condition, and extent of cause evaluations appropriately considered the safety culture components.

f. Findings

No findings were identified.

2.03 Corrective Actions a. As directed by IP 95001, determine that (1) the licensee specified appropriate corrective actions for each root and/or contributing cause, or (2) an evaluation that states no actions are necessary is adequate.

The root cause, apparent cause, and collective root cause reports identified appropriate corrective actions to address the root, contributing, and common causes for the individual reactor scrams and collective performance issues. The inspectors determined that corrective actions for the reactor scrams and common cause evaluation were reasonable, with specific actions to address the personnel, procedural, and equipment issues associated with the white PI and its associated individual reactor scram inputs.

b. As directed by IP 95001, determine that the licensee prioritized corrective actions with consideration of risk significance and regulatory compliance.

The inspectors noted that immediate corrective actions for each of the reactor scrams were performed in a timely manner to support plant restart. Longer term actions were scheduled in an appropriate time frame. Overall the inspectors determined that the corrective actions were prioritized commensurate with their significance.

c. As directed by IP 95001, determine that the licensee established a schedule for implementing and completing the corrective actions.

Corrective actions to prevent recurrence, as well as a significant number of lower-tier corrective and preventive actions, identified in the root cause and apparent cause evaluations had been completed at the time of this inspection.

A change management plan for implementing a site wide risk process (Exelon procedure AD-AA-3000) was still open at the time of this inspection. This change management plan was nearly complete with only a few remaining procedures to be revised. The inspectors met with the responsible manager of the change management plan and determined that the remaining procedure revisions would be complete by the end of 2014. The inspectors determined that the due dates for the open actions were reasonable. d. As directed by IP 95001, determine that the licensee developed quantitative and/or qualitative measures of success for determining the effectiveness of the corrective actions to preclude repetition. Effectiveness reviews for the root and apparent cause evaluations were assigned but not completed at the time of the inspection. The inspectors verified that the due dates for these effectiveness reviews were reasonable. However, the inspectors identified one instance in which an effectiveness review was not assigned or completed. Exelon did not assign an effectiveness review for the apparent cause evaluation associated with the July 11, 2014 reactor scram. Exelon procedure PI-AA-125-an effectiveness review be completed for all significance level 1 and 2 apparent cause evaluations; however, one was not completed for the July 11, 2014 event. The inspectors determined that this was a performance deficiency but screened it as minor since the corrective actions taken to date to correct the condenser bellows and second stage reheater valve have proven to be effective and have not resulted in any further plant transients. Exelon entered this issue into the corrective action program (IR 2423191). e. actions adequately address the Notice of Violation (NOV) that was the basis for the supplemental inspection, if applicable. The NRC staff did not issue an NOV to the licensee; therefore, this inspection requirement was not applicable.

f. Findings

No findings were identified. 2.04 Evaluation of IMC 0305 Criteria for Treatment of Old Design Issues The inspectors determined this issue did not meet the IMC 0305 criteria for treatment as an old design issue.

4OA6 Exit Meeting

On December 11, 2014, the inspectors presented the inspection results to Mr. J. Dostal, Plant Manager, and other members of his staff. The inspectors asked Exelon if any of the material examined during the inspection should be considered proprietary. Exelon did not identify any proprietary information. Regulatory Performance Meeting Following the December 11, 2014 exit meeting, the NRC discussed with Exelon its performance at Oyster Creek in accordance with IMC 0305, Section 10.01.a. The meeting was attended by the Region I Division of Reactor Projects, Projects Branch 6, Branch Chief, NRC inspectors, the Oyster Creek Plant Manager, and other Exelon staff. During this meeting, the NRC and Exelon discussed the issues related to the white PI for unplanned scrams that resulted in Oyster Creek being placed in the Regulatory Response Column of the Action Matrix. This discussion included the causes, corrective actions, extent of condition and extent of cause for the issues identified as a result of the white PI.

ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

G. Stathes, Site Vice President

J. Dostal, Plant Manager

M. Arnao, Operations Services Manager

A. Bready, Site Risk Analyst

J. Clark, Engineering Programs Manager

F. Jordan, Risk Classification Manager

M. McKenna, Licensing Manager

W. Saraceno, ERT Branch Manager

R. Smith, NSSS Branch Manager

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened/Closed

05000219/2014010-01 FIN Failure to Evaluate a Temporary Configuration Change (Section 4OA3)

Closed

05000219/LER-2013-004-00 LER Manual Scram due to Rise in Reactor Pressure during Turbine Valve Testing (Section 4OA3)

05000219/LER-2014-001-00 LER Manual Scram due to Lowering Vacuum (Section 4OA3)

LIST OF DOCUMENTS REVIEWED

Condition Reports

1567196

1568503

1597041

1597572

1605893

1680755

1681201

1687264

2385412

2390842

2400821

2400831

2422831*

2423121*

2423191* *Issued as a result of NRC inspection.

Attachment Procedures 2400-GMM-3900.52, Inspection and Torquing of Bolted Connection, Revision 7 2400-SMM-3411.26, Turbine Control Valve Hydraulic Enclosure, Revision 9

AD-AA-3000, Nuclear Risk Management Process, Revision 0

CC-AA-102, Design Input and Configuration Change Impact Screening, Revision 28

CC-AA-103, Configuration Change Control for Permanent Physical Plant Changes, Revision 25

CC-AA-103-1003, Owner's Acceptance Review of External Technical Products, Revision 11

CC-AA-106-1001, Configuration Change Walkdowns, Revision 5

CC-AA-103-1001, Configuration Change Control Guidance, Revision 5

CC-AA-107-1001, Post Modification Acceptance Testing, Revision 5

CC-MA-102-1001, Design Inputs and Impact Screening - Implementation, Revision 11

CC-AA-112, Temporary Configuration Changes, Revision 20

CC-AA-404, Maintenance Specification: Application Selection, Evaluation and Control of Temporary Leak Repairs, Revision 8

ER-AA-200, Preventive Maintenance Program, Revision 0

ER-AA-600-1015, FPIE PRA Model Update, Revision 17