Text

IR 05000237/2014004, 05000249/2014004; on 07/01/2014 - 09/30/2014; Dresden Nuclear Power Station, Units 2 & 3; Operability Determinations and Functional Assessments, Maintenance of Emergency Preparedness
	ML14296A594
Person / Time
Site:	Dresden
Issue date:	10/23/2014
From:	Jamnes Cameron; Reactor Projects Region 3 Branch 4
To:	Pacilio M; Exelon Generation Co, Exelon Nuclear
References
	IR 2014004
	Download: ML14296A594 (45)
	v • d • e

UNITED STATES ber 23, 2014

SUBJECT:

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 INTEGRATED INSPECTION REPORT 05000237/2014004; 05000249/2014004

Dear Mr. Pacilio:

On September 30, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Dresden Nuclear Power Station, Units 2 and 3. The enclosed report documents the results of this inspection, which were discussed on October 9, 2014, with Mr. S. Marik, and other members of your staff.

NRC inspectors documented two findings of very low safety significance (Green) in this report.

Both of these findings involved violations of NRC requirements; one of these violations was determined to be Severity Level IV under the traditional enforcement process. Further, the inspectors documented a licensee-identified violation which was determined to be of very low safety significance and is listed in Section 4OA7 in this report. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the subject or severity of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission-Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Dresden Nuclear Power Station. In addition, if you disagree with the cross-cutting aspect assigned to any finding 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 III, and the NRC Resident Inspector at the Dresden Nuclear Power Station.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and 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 System (PARS)

M. Pacilio-2-component of NRC's Agencywide Documents Access and Management System (ADAMS),

accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Jamnes L. Cameron, Chief Branch 4 Division of Reactor Projects Docket Nos. 50-237; 50-249 License Nos. DPR-19; DPR-25

Enclosure:

IR 05000237/2014004; 05000249/2014004 w/Attachment: Supplemental Information

REGION III==

Docket Nos: 05000237; 05000249 License Nos: DPR-19; DPR-25 Report No: 05000237/2014004; 05000249/2014004 Licensee: Exelon Generation Company, LLC Facility: Dresden Nuclear Power Station, Units 2 and 3 Location: Morris, IL Dates: July 1, 2014 through September 30, 2014 Inspectors: G. Roach, Senior Resident Inspector D. Lords, Resident Inspector S. Barr, Senior Emergency Preparedness Inspector, Region 1 J. Beavers, Emergency Preparedness Inspector J. Kutlesa, Physical Security Inspector J. Laughlin, Emergency Preparedness Specialist, NSIR T. Smith, Emergency Preparedness Specialist, NSIR Approved by: J. Cameron, Chief Projects Branch 4 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

Inspection Report 05000237/2014004, 05000249/2014004; 07/01/2014-09/30/2014; Dresden

Nuclear Power Station, Units 2 & 3; Operability Determinations and Functional Assessments,

Maintenance of Emergency Preparedness.

This report covers a 3-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. Two Green findings were identified by the inspectors. The findings were considered non-cited violations (NCV) of NRC regulations. The significance of inspection findings is indicated by their color (i.e., greater than Green, or Green,

White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process dated June 2, 2011. Cross cutting aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas effective date January 1, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated July 9, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process Revision 5, dated February 201

NRC-Identified Findings

Cornerstone: Mitigating Systems

Severity Level IV. The inspectors identified a NCV of 10 CFR 50.59, Changes, Tests and Experiments, when, on February 10, 2011, the licensee failed to complete a 10 CFR 50.59 evaluation when they revised procedure DOP 1300-02 to change the position of Motor Operated Valve (MOV) 2-1301-3, Reactor Inlet Isolation, such that the Isolation Condenser (IC) system would not meet its design requirement of removing 84.2E+06 BTUs in 20 minutes when initiated from its minimum Technical Specification (TS) level and maximum TS temperature.

The inspectors determined that the licensees failure to identify that the valve position adjustment required a 10 CFR 50.59 evaluation was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. This finding was more than minor because there was a reasonable likelihood that the change would have required NRC review and approval prior to implementation. Specifically, by establishing a new position setting of MOV 2-1301-3, the licensee failed to determine that the proposed change would cause isolation condenser tubes to become exposed in the design basis accident such that it adversely affected a Final Safety Analysis Report described design function, which required an evaluation to be performed. In accordance with IMC 0612, Appendix B, Issue Screening, traditional enforcement does apply as the violation impacted the regulatory process. Using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012,

Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of the system and/or function, did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time, and did not result in the actual loss of one or more trains of non-technical specification equipment.

Inspectors assessed the violation in accordance with the Enforcement Policy, and determined it to be a Severity Level IV violation because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). This finding has a cross-cutting aspect of Design Margins

[IMC 0310, H.6] in the area of human performance, for failing to carefully guard and maintain the IC design requirement of removing 84.2E+06 BTU in 20 minutes.

(Section 1R15.b.1)

Cornerstone: Emergency Preparedness

Green.

The NRC identified a NCV of 10 CFR 50.54(q)(2) associated with 10 CFR 50.47(b)(10) and 10 CFR Part 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Dresden Nuclear Power Station Emergency Plan as a result of failing to provide the station evacuation time estimate (ETE) to the responsible offsite response organizations (OROs) by the required date.

Exelon submitted the Dresden Nuclear Power Station ETE to the NRC on December 12, 2012, prior to the required due date of December 22, 2012. The NRC completeness review found the ETEs to be incomplete due to Exelon fleet common and site-specific deficiencies, thereby preventing Exelon from providing the ETEs to responsible OROs and from updating site-specific protective action strategies as necessary. The NRC discussed its concerns regarding the completeness of the ETE, in a teleconference with Exelon on June 10, 2013, and on September 5, 2013, Exelon resubmitted the ETEs for its sites. The NRC again found the ETEs to be incomplete.

The issue is a performance deficiency because it involves a failure to comply with a regulation that was under Exelons control to identify and prevent. The finding is more than minor because it is associated with the emergency preparedness cornerstone attribute of procedure quality and because it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency.

The finding is of very low safety significance because it was a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The licensee had entered this issue into their corrective action program (CAP) and re-submitted a new revision of the Dresden Nuclear Power Station ETE to the NRC on May 2, 2014, which was found to be complete by the NRC. The cause of the finding is related to the cross-cutting element of Human Performance, Documentation. [IMC 0310, H.7] (Section 1EP5.2.b)

Licensee-Identified Violations

Violations of very low safety significance or Severity Level IV that were identified by the licensee have been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensees corrective CAP. These violations and CAP tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 2 Unit 2 began the inspection period near full power. On July 5, 2014, operators reduced power to approximately 25 percent for maintenance-related drywell entries, and returned to full power on July 6th. On September 6, 2014, operators reduced power to approximately 25 percent for maintenance-related drywell entries, and returned to full power on September 8th. The unit operated at or near full power for the remainder of the inspection period.

Unit 3 Unit 3 operated at or near full power capability for the duration of the inspection period, with brief down power evolutions to perform control rod sequence exchanges. On September 1, 2014, the operators entered coastdown in preparation for refueling outage Dresden Unit 3 Refuel Outage 23 (D3R23) which is scheduled to begin on November 3,

REACTOR SAFETY

Cornerstone: Initiating Events, Mitigating Systems, Barrier Integrity, and

Emergency Preparedness

1R01 Adverse Weather Protection

Main article: IP 71111.01

.1 External Flooding

a. Inspection Scope

The inspectors evaluated the design, material condition, and procedures for coping with the design basis probable maximum flood. The evaluation included a review to check for deviations from the descriptions provided in the Updated Final Safety Analysis Report (UFSAR) for features intended to mitigate the potential for flooding from external factors.

As part of this evaluation, the inspectors checked for obstructions that could prevent draining, checked that the roofs did not contain obvious loose items that could clog drains in the event of heavy precipitation, and determined that barriers required to mitigate the flood were in place and operable. Additionally, the inspectors performed a walkdown of the protected area to identify any modification to the site which would inhibit site drainage during a probable maximum precipitation event or allow water ingress past a barrier. The inspectors also walked down underground bunkers/manholes subject to flooding that contained multiple train or multiple function risk-significant cables. The inspectors also reviewed the abnormal operating procedure (AOP) for mitigating the design basis flood to ensure it could be implemented as written and that equipment identified in the procedure was properly staged, maintained and tested as appropriate.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted one external flooding sample as defined in Inspection Procedure (IP) 71111.01-05.

b. Findings

No findings were identified.

1R04 Equipment Alignment

Main article: IP 71111.04

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant systems:

2A fuel pool cooling system with 2B out of service (OOS);

3A core spray (CS) with 3B CS OOS;

2/3B standby gas treatment system (SBGT) with 2/3A SGBT OOS; and

2/3B control room emergency ventilation system (CREV) upon return to service.

The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, UFSAR, TS requirements, outstanding work orders (WOs), condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the CAP with the appropriate significance characterization. Documents reviewed are listed in the to this report.

These activities constituted four partial system walkdown samples as defined in IP 71111.04-05.

b. Findings

No findings were identified.

.2 Semi-Annual Complete System Walkdown

a. Inspection Scope

On July 17-21, 2014, the inspectors performed a complete system alignment inspection of the Unit 3 high pressure coolant injection system (HPCI) to verify the functional capability of the system. This system was selected because it was considered both safety significant and risk significant in the licensees probabilistic risk assessment. The inspectors walked down the system to review mechanical and electrical equipment lineups; electrical power availability; system pressure and temperature indications, as appropriate; component labeling; component lubrication; component and equipment cooling; hangers and supports; operability of support systems; and to ensure that ancillary equipment or debris did not interfere with equipment operation. A review of a sample of past and outstanding WOs was performed to determine whether any deficiencies significantly affected the system function. In addition, the inspectors reviewed the CAP database to ensure that system equipment alignment problems were being identified and appropriately resolved. Documents reviewed are listed in the to this report.

These activities constituted one complete system walkdown sample as defined in IP 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

Main article: IP 71111.05

.1 Routine Resident Inspector Tours

Main article: IP 71111.05Q

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

U2/3 off gas filter building, elevation 476, Fire Zone 14.4;

U3 turbine area, elevation 561, Fire Zone 8.2.8A;

U3 secondary containment, elevation 570, Fire Zone 1.1.1.4; and

U2 secondary containment, elevation 570, Fire Zone 1.1.2.4, The inspectors reviewed areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and implemented adequate compensatory measures for out-of-service, degraded or inoperable fire protection equipment, systems, or features in accordance with the licensees fire plan. The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to impact equipment which could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event.

Using the documents listed in the Attachment to this report, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees CAP. Documents reviewed are listed in the Attachment to this report.

These activities constituted four quarterly fire protection inspection samples as defined in IP 71111.05-05.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

Main article: IP 71111.11

.1 Resident Inspector Quarterly Review of Licensed Operator Requalification

Main article: IP 71111.11Q

a. Inspection Scope

On August 11, 2014 and August 25, 2014, the inspectors observed a crew of licensed operators in the plants simulator during licensed operator requalification training to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

licensed operator performance;

crews clarity and formality of communications;

ability to take timely actions in the conservative direction;

prioritization, interpretation, and verification of annunciator alarms;

correct use and implementation of abnormal and emergency procedures;

control board manipulations;

oversight and direction from supervisors; and

ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications.

The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements. Documents reviewed are listed in the Attachment to this report.

This inspection constituted two quarterly licensed operator requalification program simulator samples as defined in IP 71111.11.

b. Findings

No findings were identified.

.2 Resident Inspector Quarterly Observation of Heightened Activity or Risk

Main article: IP 71111.11Q

a. Inspection Scope

On September 7, 2014, the inspectors observed control room operators perform a power ascension to full power following a down power to 25 percent reactor power in order to support a containment entry to add oil and modify the oil reservoir system of the 2A reactor recirculation pump. This was an activity that required heightened awareness or was related to increased risk. The inspectors evaluated the following areas:

licensed operator performance;

crews clarity and formality of communications;

ability to take timely actions in the conservative direction;

prioritization, interpretation, and verification of annunciator alarms (if applicable);

correct use and implementation of procedures;

control board (or equipment) manipulations;

oversight and direction from supervisors; and

ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications (if applicable).

The performance in these areas was compared to pre-established operator action expectations, procedural compliance and task completion requirements. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one quarterly licensed operator heightened activity/risk sample as defined in IP 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

Main article: IP 71111.12

.1 Routine Quarterly Evaluations

Main article: IP 71111.12Q

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk-significant systems:

Unit 2/3 miscellaneous drains and sumps;

Unit 3 battery room heating, ventilation and air conditioning; and

secondary containment interlock doors.

The inspectors reviewed events such as where ineffective equipment maintenance had or could have resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

implementing appropriate work practices;

identifying and addressing common cause failures;

scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;

characterizing system reliability issues for performance;

charging unavailability for performance;

trending key parameters for condition monitoring;

ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and

verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2), or appropriate and adequate goals and corrective actions for systems classified as (a)(1).

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

This inspection constituted three quarterly maintenance effectiveness samples as defined in IP 71111.12-05.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

Main article: IP 71111.13

.1 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed the licensee's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

Unit 2 YELLOW risk with 2B fuel pool cooling out of service;

Unit 2 rod block monitor-no rod selected light out;

emergent work on Unit 3 core spray valve 3-1402-24B;

abnormal excitation spikes on the Unit 2 digital automatic voltage regulator; and

Unit 2 emergency diesel generator troubleshooting and emergent repairs following fast start failure.

These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Documents reviewed during this inspection are listed in the Attachment to this report.

These maintenance risk assessments and emergent work control activities constituted five samples as defined in IP 71111.13-05.

b. Findings

No findings were identified.

1R15 Operability Determinations and Functional Assessments

Main article: IP 71111.15

.1 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

Unit 2 isolation condenser excessive moisture carryover;

Oyster Creek Nuclear Plant Operating Experience regarding failed electromagnetic relief valve solenoid actuator;

Unit 3 125 Vdc battery system experiencing hard ground;

seismic concerns due to structural supports in the vicinity of the Unit 2 and Unit 3 torus;

failure to capture Unit 2 emergency diesel generator starting criteria during fast start surveillance; and

containment cooling service water (CCSW) flow meter stuck during CREV operability surveillance.

The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and UFSAR to the licensees evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Documents reviewed are listed in the to this report.

This operability inspection constituted six samples as defined in IP 71111.15-05.

b. Findings

(1) Failure to Perform Adequate 10 CFR 50.59 Evaluation For DOP 1300-02

Introduction:

The inspectors identified a Severity Level IV NCV of 10 CFR 50.59, Changes, Tests, and Experiments, and an associated Green finding for the licensees failure to perform a written evaluation of the procedure change to establish the position of MOV 2-1301-3. Specifically, on February 10, 2011, as part of Revision 24 to DOP 1300-02, Automatic Operation of Isolation Condenser (IC), the licensee incorrectly approved the 50.59 screen to determine whether the proposed change to the position setting of MOV 2-1301-3 involved a change to structures, systems, or components (SSC) such that it adversely affected a Final Safety Analysis Report described design function.

Description:

On February 10, 2011, Dresden Station approved procedure DOP 1300-02, Automatic Operation of Isolation Condenser, Revision 24, which established the new position setting of MOV 2-1301-3, Reactor Inlet Isolation. On February 8, 2011, the licensee performed a 10 CFR 50.59 Applicability Review for the procedure change. The review states: This activity ensures the SSC meets design limits and is thus a maintenance activity. Therefore this activity is not subject to review under 10 CFR 50.59.

On December 19, 2013, following the next performance of the 5-year heat capacity test for the Unit 2 Isolation Condenser, the licensee wrote IR 1599386 which states Because of the higher carryover measured in 2013, the Unit 2 Isolation Condenser cannot meet its design requirement of removing 84.2E+06 BTUs in 20 minutes. The licensee then performed Operability Evaluation EC 396493, 2-1302 Isolation Condenser, Revision 1 which states that the Unit 2 Isolation Condenser cannot meet its design requirement of removing 84.2E+06 BTUs in 20 minutes (or in any amount of time), and also that the heat removal rate is governed by the open position of the 2-1301-3 valve, which is not expected to change with time. Nuclear Energy Institute (NEI) 96-07, Revision 1, Guidelines for 10 CFR 50.59 Implementation, dated November 2000 (endorsed by the NRC in Regulatory Guide 1.187) defines, on page 20, a malfunction of SSCs important to safety as the failure of an SSC to perform its intended design function described in the UFSAR. This guidance also states that a change to a structure, system, or component, such that it adversely affects a Final Safety Analysis Report described function is a change that is controlled by 10 CFR Part 50.59.

The Technical Specification bases for the values in SR 3.5.3.1 were documented in calculation BSA D-95-07, Dresden Isolation Condenser Performance. This calculation, which was based on a starting water level and temperature in the shell of 6 feet and 210 degrees Fahrenheit (°F), concluded the IC would be able to meet its required heat removal capability. However, the calculation noted in order to accomplish the required heat removal capability, the liquid water level in the shell would drop to approximately the last quarter of the tube bundle (26.3 percent tube bundle height) after 20 minutes of operation, exposing 3/4 of the tube bundles. Based upon results of the U2 isolation condenser heat capacity test performed on November 5, 2013, as documented in IR 1599386, the amount of water carryover during the test would result in reaching the mid-plane of the tube bundles in 10.25 minutes. Therefore, because of the higher carryover measured after the position adjustment of MOV 2-1301-3 made in 2008, the Unit 2 isolation condenser could not meet its design requirement of removing 84.2E+06 BTUs in 20 minutes (or in any amount of time).

Analysis:

The inspectors determined that the licensees failure to perform 10 CFR 50.59 evaluation prior to implementing its new procedure was a licensee performance deficiency warranting a significance evaluation. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. This finding was more than minor because there was a reasonable likelihood that the change would have required NRC review and approval prior to implementation. Specifically, by establishing a new position setting of MOV 2-1301-3, the licensee failed to determine that the proposed change would cause isolation condenser tubes to become exposed in the design basis accident such that it resulted in a more than minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the Final Safety Analysis Report, which required an evaluation to be performed. Using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of the system and/or function, did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time, and did not result in the actual loss of one or more trains of non-technical specification equipment.

Inspectors assessed the violation in accordance with the Enforcement Policy, and determined it to be a Severity Level IV violation because it resulted in a condition evaluated by the Significance Determination Process as having very low safety significance (Enforcement Policy example 6.1.d.2).

This finding has a cross-cutting aspect of Design Margins (H.6) in the area of human performance, for failing to carefully guard and maintain the IC design requirement of removing 84.2E+06 BTU in 20 minutes. Specifically, the licensee did not appropriately apply the 50.59 screening process when it concluded that the position adjustment of MOV 2-1301-3 was a maintenance activity which could not adversely impact the design function of the IC system.

Enforcement:

Title 10 CFR 50.59(d)(1) requires licensees to maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c). These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to paragraph (c)(2) of this section.

10 CFR(c)(2)(ii) states, in part, that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the final safety analysis report(as updated). Improved Technical Specification Basis for Surveillance Requirement (SR) 3.5.3.1 states that the SR verifies the water volume and temperature in the shell side of the IC provides sufficient decay heat removal capability for 20 minutes of operation without makeup water.

Contrary to the above, on February 10, 2011, the licensee failed to maintain a record of a written evaluation for the change in position of MOV 2-1301-1. Specifically, the licensee approved and implemented procedure DOP 1300-02, Automatic Operation of Isolation Condenser, Revision 24 which altered the position of MOV 2-1301-3 such that the isolation condenser system would not meet its design requirement of removing 84.2E+06 BTUs in 20 minutes when initiated from its minimum TS level and maximum Technical Specification temperature. Because this violation was of very low safety significance, was not repetitive, did not appear to have any willful aspects, and was entered into the licensees corrective action program, this violation is being treated as a NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV/FIN 05000237/2014004-01, Failure to Perform 10 CFR 50.59 Evaluation for Procedure DOP 1300-02) This violation was entered into the licensees corrective action program as IR 1670444.

1R18 Plant Modifications

Main article: IP 71111.18

.1 Plant Modifications

a. Inspection Scope

The inspectors reviewed the following modification(s):

EC [Engineering Change] 388140, Re-Route Condensate Pipe Lines 2/3-3327-12 & 2/3-3329-16.

The inspectors reviewed the configuration changes and associated 10 CFR 50.59 safety evaluation screening against the design basis, the UFSAR, and the TS, as applicable, to verify that the modification did not affect the operability or availability of the affected system(s). The inspectors, as applicable, observed ongoing and completed work activities to ensure that the modifications were installed as directed and consistent with the design control documents; the modifications operated as expected; post-modification testing adequately demonstrated continued system operability, availability, and reliability; and that operation of the modifications did not impact the operability of any interfacing systems. As applicable, the inspectors verified that relevant procedure, design, and licensing documents were properly updated. Lastly, the inspectors discussed the plant modification with operations, engineering, and training personnel to ensure that the individuals were aware of how the operation with the plant modification in place could impact overall plant performance. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one permanent plant modification sample as defined in IP 71111.18-05.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

Main article: IP 71111.19

.1 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the following post-maintenance (PM) activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

WO 1532813, Unit 3 emergency core cooling system jockey pump post maintenance test (PMT) following pump greasing and repack;

WO 1625648, Operations PMT Support Screen Refuse Pit Upgrade-EC 392950;

WO 1102882, Drywell equipment drain system AOV 2-2001-5 following 6-year solenoid replacement;

3B core spray pump following a preventative maintenance work window; and

WO 1764314, 2/3B control room emergency ventilation train PMT following a preventative maintenance work window.

These activities were selected based upon the structure, system, or component's ability to impact risk. The inspectors evaluated these activities for the following (as applicable):

the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion); and test documentation was properly evaluated. The inspectors evaluated the activities against TSs, the UFSAR, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with post-maintenance tests to determine whether the licensee was identifying problems and entering them in the CAP and that the problems were being corrected commensurate with their importance to safety. Documents reviewed are listed in the Attachment to this report.

This inspection constituted five post-maintenance testing samples as defined in IP 71111.19-05.

b. Findings

No findings were identified.

1R20 Outage Activities

Main article: IP 71111.20

.1 Other Outage Activities

a. Inspection Scope

On September 7, 2014, the inspectors performed a closeout walk down of the Unit 2 drywell following licensee maintenance activities to install an additional oil reservoir for the 2A reactor recirculation pump which was experiencing leakage from the lower bearing oil system.

The inspectors also observed power ascension activities in the main control room and inside the plant following the brief maintenance outage, and reviewed the licensees identification and resolution of problems associated with the outage.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted one other outage sample as defined in IP 71111.20-05.

b. Findings

No findings were identified.

1R22 Surveillance Testing

Main article: IP 71111.22

.1 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements:

WO 1737669, Operations D3 Quarterly Technical Specification (TS) Emergency Diesel Generator Cooling Water Pump Test for IST [inservice testing] Program Surveillance (IST);

WO 1734455, D3 Quarterly TS Core Spray Pump Test with Torus Available for IST Data Surveillance (Routine);

WO 1737283, D2 Quarterly TS Core Spray Pump Test with Torus Available for IST Data Surveillance (Routine);

WO 1716647, Operations Dresden 2 Semi Annual TS Diesel Generator Fast Start Operability Surveillance (Routine);

WO 1742981, Operations D3 Quarterly TS Main Steam Isolation Valve Scram Functional Test (Routine); and

WO 1703750, Unit 2 Isolation Condenser Five Year Heat Removal Capability Test (Routine).

The inspectors observed in-plant activities and reviewed procedures and associated records to determine the following:

did preconditioning occur;

the effects of the testing were adequately addressed by control room personnel or engineers prior to the commencement of the testing;

acceptance criteria were clearly stated, demonstrated operational readiness, and were consistent with the system design basis;

plant equipment calibration was correct, accurate, and properly documented;

as-left setpoints were within required ranges; and the calibration frequency was in accordance with TSs, the USAR, procedures, and applicable commitments;

measuring and test equipment calibration was current;

test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied;

test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used;

test data and results were accurate, complete, within limits, and valid;

test equipment was removed after testing;

where applicable for inservice testing activities, testing was performed in accordance with the applicable version of Section XI, American Society of Mechanical Engineers code, and reference values were consistent with the system design basis;

where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable;

prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test;

equipment was returned to a position or status required to support the performance of its safety functions; and

all problems identified during the testing were appropriately documented and dispositioned in the CAP.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted five routine surveillance testing samples and one inservice testing sample. as defined in IP 71111.22, Sections-02 and-05.

b. Findings

No findings were identified.

1EP2 Alert and Notification System Evaluation

Main article: IP 71114.02

a. Inspection Scope

The inspectors reviewed a sample of nuclear oversight staffs audits of the EP Program to determine whether these independent assessments met the requirements of 10 CFR 50.54(t). The inspectors also reviewed critique reports and samples of CAP records associated with the 2013 Biennial Exercise, as well as various EP drills conducted, in order to determine that the licensee fulfilled its drill commitments and to evaluate the licensees efforts to identify, track, and resolve concerns identified during these activities. The inspectors reviewed a sample of EP items and corrective actions related to the facilitys EP Program and activities from February 2012 through July 2014 to determine whether corrective actions were completed in accordance with the site's CAP. Documents reviewed are listed in the Attachment to this report.

This correction of EP weaknesses and deficiencies inspection constituted one sample as defined in IP 71114.05-06.

b. Findings

No findings were identified.

.2 NRC Review of Licensees Evacuation Time Estimate

a. Inspection Scope

Nuclear Regulatory Commission EP rulemaking, which became effective on December 23, 2011, added a new regulation that required a licensee to develop an evacuation time estimate (ETE) analysis and submit it to the NRC by December 22, 2012. This inspection was a follow-up of issues identified by the NRC headquarters staff during its review of the Exelon submittal of the ETE for the ten sites that it operates.

The NRC headquarters staff related those issues to Exelon, which provided responses through 2013 and into 2014. During this inspection period, regional EP inspectors reviewed applicable licensee documents, conducted discussions with licensee personnel, and provided assessment of the Exelon response.

This emergency preparedness inspection constituted no samples as defined in IP 71114.05.

b. Findings

Introduction:

The NRC identified a Green NCV of 10 CFR 50.54(q)(2) for failing to maintain the effectiveness of the Dresden Nuclear Power Station Emergency Plan.

Specifically, the licensee failed to provide the station ETE to responsible OROs and failed to update their site-specific protective action strategies as necessary as required by 10 CFR 50.47(b)(10), and Section IV, Paragraph 4 of Appendix E to 10 CFR Part 50.

Description:

The NRC issued final new and amended emergency preparedness regulations on November 23, 2011 (76 Federal Register 72560). This rulemaking, which became effective on December 23, 2011, amended 10 CFR 50.47(b)(10) to require licensees to update the ETE on a periodic basis. The rulemaking also added a new regulation 10 CFR Part 50, Appendix E, Section IV.4, which requires a licensee to develop an ETE analysis using the most recent decennial census data and submit it to the NRC within 365 days of December 23, 2011. Concurrently with the issuance of the rulemaking, the NRC published a new report entitled Criteria for Development of Evacuation Time Estimate Studies, NUREG/CR-7002. The Statements of Consideration for the rulemaking (76 Federal Register 72580) identified that the NRC would review the submitted ETEs for completeness using that document. The Statements also provided that the guidance of NUREG/CR-7002 guidance was an acceptable template to meet the requirements and that licensees should use the guidance or an appropriate alternative.

By individual letters dated December 12, 2012, Exelon submitted the ETEs for the sites for which it holds the operating licenses, including Dresden Nuclear Power Station. By a letter dated January 23, 2013, Exelon submitted the NUREG/CR-7002 checklists for these ETEs. These checklists identified where a particular criterion was addressed in the ETEs, facilitating the NRC review.

As provided in the Statements of Consideration, the NRC performed a completeness review using the checklists and found the ETEs (including that for the Dresden Nuclear Power Station) to be incomplete due to common and site-specific deficiencies.

The NRC discussed its concerns regarding the completeness of the ETEs, in a teleconference with Exelon conducted on June 10, 2013. By letter dated September 5, 2013, Exelon resubmitted the ETEs and the associated checklists for its sites. The NRC performed another completeness review and again found the ETEs to be incomplete.

Examples of information missing from the submittal included: peak and average attendance were not stated (NUREG/CR-7002 Criteria Item 2.1.2.a); the ETE used a value based on campsite and hotel capacity, vice an average value (2.1.2.b); basis for speed and capacity reduction factors due to weather was not provided (3.4.b); snow removal was not addressed (3.4.c); no bus routes or plans were included in the ETE analysis (4.1.2.a); and, no discussion on the means of evacuating ambulatory and non-ambulatory residents was included (4.1.2.b).

Exelon entered this issue into their CAP as IR1578649. Exelon submitted a third ETE for Dresden Nuclear Power Station on May 2, 2014, and the NRCs review of that ETE was found complete and documented in Section 1EP4 of this report.

Analysis:

The inspectors determined that Exelons failure to submit a complete updated ETE for the Dresden Nuclear Power Station by December 22, 2012, was a licensee performance deficiency because the issue was a failure to comply with a regulatory requirement and the issue was reasonably within the licensees ability to foresee and correct, and therefore should have been prevented, for both the December 12, 2012, and September 5, 2013, submittals.

Using IMC 0612, Appendix B, Issue Screening, the inspector determined that the performance deficiency is associated with the emergency preparedness cornerstone attribute of procedure quality and was more than minor because it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The ETE is an input into the development of protective action strategies prior to an accident and to the protective action recommendation decision making process during an accident. Inadequate ETEs have the potential to reduce the effectiveness of public protective actions implemented by the OROs.

The inspectors utilized IMC 0609, Appendix B, Emergency Preparedness (EP)

Significance Determination Process (SDP), to determine the significance of the performance deficiency. The performance deficiency was associated with planning standard 10 CFR 50.47(b)(10). EP SDP Table 5.10-1, Significance Examples

§50.47(b)(10), provides two Green significance examples: ETEs and updates to the ETEs were not provided to responsible OROs, and The current public protective action strategies documented in emergency preparedness implementing procedures (EPIPs)are not consistent with the current ETE. The inspector concluded that, because the performance deficiency delayed the NRCs approval of the Dresden Nuclear Power Station ETE, the ETE was not provided to the site OROs nor was it used to inform the site EPIPs as required by 10 CFR 50.47(b)(10), and Section IV, Paragraph 4 of Appendix E to 10 CFR Part 50. Therefore, In accordance with EP SDP Table 5.10-1, this finding screened as a Green finding.

This finding had a cross-cutting aspect in the area of Human Performance, Documentation, because Exelon personnel did not create and maintain complete, accurate and, up-to-date documentation. Specifically, the Emergency Preparedness organization did not develop the Dresden Nuclear Power Station ETE as required by the new regulation introduced by the NRCs EP Rule. (H.7)

Enforcement:

Requirements in 10 CFR 50.54(q)(2) state, in part, that a licensee shall follow and maintain in effective emergency plans which meet the standards in 10 CFR 50.47(b) and the requirements in Appendix E to this part. Title 10 CFR 50.47(b)(10), requires, in part, that licensees shall develop an evacuation time estimate and update it on a periodic basis. Title 10 CFR Part 50 Appendix E, Section IV.4, states that within 365 days of December 23, 2011, nuclear power reactor licensees shall develop an ETE analysis and submit it under § 50.4.

Contrary to the above, within 365 days of December 23, 2011, Exelon, the licensee for Dresden Nuclear Power Station, failed to develop a complete and adequate ETE analysis and submit it under 10 CFR 50.4. Immediate corrective actions taken by Exelon included entering this issue into their CAP as issue report (IR) 1578649 and revising the ETE to satisfy NRC requirements. Because this finding is of very low safety significance (Green) and was entered into Exelons CAP, this issue is being treated as an NCV consistent with Section 2.3.2.a. of the Enforcement Policy.

(NCV 05000237/2014004-02; 05000249/2014004-02: Inadequate Evacuation Time Estimate Submittals).

1EP6 Drill Evaluation

Main article: IP 71114.06

.1 Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of routine licensee emergency response organization training on July 22, 2014, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed emergency response operations in the main control room simulator to determine whether the event classification, notifications, and protective action recommendations were performed in accordance with procedures.

The inspectors also attended the licensee post-training critique to compare any inspector-observed weakness with those identified by the licensee staff in order to evaluate the critique and to verify whether the licensee staff was properly identifying weaknesses and entering them into the CAP. As part of the inspection, the inspectors reviewed the drill package and other documents listed in the Attachment to this report.

This emergency preparedness training event inspection constituted one sample as defined in IP 71114.06-06.

b. Findings

No findings were identified.

OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical Protection

4OA1 Performance Indicator Verification

Main article: IP 71151

.1 Mitigating Systems Performance IndexHeat Removal System

.6 Alert and Notification System

a. Inspection Scope

The inspectors sampled licensee submittals for the alert and notification system (ANS)

Reliability PI (EP03) for the period from the third quarter 2013 through the second quarter 2014. The inspectors used PI definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 2013, to determine the accuracy of the PI data reported during those periods. The inspectors reviewed the licensees records and processes including procedural guidance on assessing opportunities for the PI and results of periodic ANS operability tests to validate the accuracy of the submittals. The inspectors also reviewed the licensees IR database to determine whether any problems had been identified with the PI data collected or transmitted for this indicator and none were identified.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted one ANS reliability sample as defined in IP 71151-05.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

Main article: IP 71152

.1 Routine Review of Items Entered into the Corrective Action Program

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify they were being entered into the licensees CAP at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Attributes reviewed included: identification of the problem was complete and accurate; timeliness was commensurate with the safety significance; evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent-of-condition reviews, and previous occurrences reviews were proper and adequate; and that the classification, prioritization, focus, and timeliness of corrective actions were commensurate with safety and sufficient to prevent recurrence of the issue.

Minor issues entered into the licensees CAP as a result of the inspectors observations are included in the Attachment to this report.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

No findings were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees CAP. This review was accomplished through inspection of the stations daily condition report packages.

These daily reviews were performed by procedure as part of the inspectors daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings were identified.

.3 Selected Issue Follow-Up Inspection: Review of Corrective Actions Associated with

NRC Identified Inadequate External Flooding Procedure

a. Inspection Scope

During Supplemental Inspection 95001, as documented in Dresden Supplemental Inspection Report 2013010 [ML accession number 14023A004], the inspectors reviewed the licensees root cause, completed corrective actions, and planned corrective actions associated with a White finding for an inadequate external flooding procedure DOA 0010-04, Floods. The White finding was originally described in Dresden Integrated Inspection Report 2013002 [ML13128A056].

The inspectors reviewed the completion of additional corrective actions with due dates after the completion of the Supplemental Inspection and noted the scheduling of an effectiveness review for the licensee identified correction actions to prevent recurrence in the spring of 2015. Specifically, the inspectors verified the development and reviewed the content of case studies covering the failure to recognize vulnerabilities in the flood strategy and the implications of site personnel possessing a minimum compliance culture. The licensees root cause for the White finding was determined to be the sites minimum compliance culture with regard to low probability, high risk events. These case studies were presented as required training for Shift Managers, regulatory assurance personnel, engineering personnel, first line supervisors, Site Ownership Committee members, Management Review Committee members, Work Week Managers, and senior site managers. A 3-year continuous training cycle covering these case studies was developed for the groups previously mentioned. The inspectors noted that there were two engineering individuals being tracked by the licensee as not having received the case study training due to being on long term medical leave. Due dates during the fourth quarter of 2014 were identified for these individuals to receive the required training upon return to duty. Documents reviewed are listed in the Attachment to this report.

This review constituted one in-depth problem identification and resolution sample as defined in IP 71152-05.

b. Findings

No findings were identified.

.4 Selected Issue Follow-Up Inspection: Review of Licensee Corrective Actions Associated

with Recent Human Performance Failures

a. Inspection Scope

The inspectors reviewed a licensee human performance improvement initiative developed by the operations department as well as site wide corrective action documents covering various events and issues which have occurred over the past year having been attributed to deficient human performance practices. In addition, the inspectors performed in-field observations of licensee personnel utilizing human performance tools in the course of their daily activities as well as attended pre-shift and pre-job briefs where human performance topics were discussed and licensee personnel were quizzed by their supervisors on the expected human performance tools they would be using to accomplish their job. In addition, the inspectors discussed with operations management the sites implementation of an Operator Excellence Program in which shift management-identified operators receive additional coaching and peer check of their activities as they develop improved operator performance. Documents reviewed are listed in the Attachment to this report.

This review constituted one in-depth problem identification and resolution sample as defined in IP 71152-05.

b. Findings

No findings were identified.

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

Main article: IP 71153

.1 (Closed) Licensee Event Report 05000237/2014-002-01; Unit 2 Reactor Scram Due to

Main Power Transformer Failure On April 12, 2014 at 1012, the licensee received a Unit 2 reactor scram due to a main power transformer (MPT) sudden pressure trip. Operators immediately ensured the plant was in a safe hot shutdown (Mode 3) condition. All plant equipment operated as expected.

At the time of submittal of this Supplemental licensee event report (LER), the vendor had not completed its Failure Analysis Report. Once the licensee has received this document it will complete a Root Cause Analysis for the event and develop corrective actions to prevent recurrence. The licensee expects to submit a final LER for this event on January 8, 2015.

This event was reported in accordance with 10 CFR 50.73(a)(2)(iv)(A), any event or condition that resulted in manual or automatic actuation of the reactor protection system.

The inspectors reviewed and closed the original event report in Dresden NRC Integrated Inspection Report 2014003. Documents reviewed are listed in the Attachment to this report. No findings or violations of NRC requirements were identified.

This LER is closed.

This event follow up review constituted one sample as defined in IP 71153-05.

.2 (Closed) Licensee Event Report 05000237/2014-003-01; Unit 2 Reactor Scram During

Automatic Voltage Regulator Channel Transfer On May 3, 2014 at 1209, the licensee received a Unit 2 reactor scram due to a main generator trip during a planned automatic voltage regulator (AVR) channel swap.

Operators immediately ensured the plant was in a safe hot shutdown (Mode 3) condition.

All plant equipment subsequently operated as expected.

At the time of submittal of this Supplemental LER, the licensee had not completed its Root Cause

Analysis.

The licensee expects to submit a final LER for this event on November 7, 2014.

This event was reported in accordance with 10 CFR 50.73(a)(2)(iv)(A), any event or condition that resulted in manual or automatic actuation of the reactor protection system.

The inspectors reviewed and closed the original event report in Dresden NRC Integrated Inspection Report 2014003. Documents reviewed are listed in the Attachment to this report. No findings or violations of NRC requirements were identified.

This LER is closed.

This event follow up review constituted one sample as defined in IP 71153-05.

4OA5 Other Activities

.1 Institute of Nuclear Power Operations Plant Assessment Report Review

a. Inspection Scope

The inspectors reviewed the final report for the Institute of Nuclear Power Operations (INPO) plant assessment conducted in June 2014. The inspectors reviewed the report to ensure that issues identified were consistent with the NRC perspectives of licensee performance and to verify if any significant safety issues were identified that required further NRC follow-up.

b. Findings

No findings were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

On October 9, 2014, the inspectors presented the inspection results to Mr. S. Marik, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

.2 Interim Exit Meetings

Interim exits were conducted for:

The results of the Emergency Preparedness Program inspection were discussed with Mr. J. Washko on August 7, 2014.

The results of the Emergency Preparedness Program inspection and ETE submittal review were discussed via teleconference with Mr. D. Doggett on September 4, 2014.

The inspectors confirmed that none of the potential report input discussed was considered proprietary. Proprietary material received during the inspection was returned to the licensee.

4OA7 Licensee-Identified Violations

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 the NRC Enforcement Policy for being disposition as a NCV.

A violation of 10 CFR 50.65(b)(2)(i) was identified by the licensee during a review of systems and components utilized in the emergency operating procedures (EOP) as compared to functions scoped into the sites Maintenance Rule Program. While reviewing emergency operating procedure DEOP 300-1, Secondary Containment Control the Site Maintenance Rule Coordinator (SMRC) noted that one of the entry criterion for the procedure included receiving a Reactor Building Floor Drain Sump (RBFDS) Hi-Hi level alarm. The SMRC noted that the RBFDS was scoped into the Maintenance Rule Program, but the Hi-Hi level alarm function was not. This event was entered into the licensees CAP as IR 1698084. The failure to scope into the Maintenance Rule Program non-safety related structures, systems, and components that are relied upon to mitigate accidents or transients or are used in plant emergency operating procedures was considered a performance deficiency. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, no maintenance performance criteria would have been established to ensure the reliability of a function serving as an entry criterion for an EOP associated with maintaining containment integrity. The inspectors evaluated the finding using IMC 0609, Significance Determination Process, 0609.04, Initial Characterization of Findings, Tables 2 and 3, and Appendix A, The Significance Determination Process (SDP) for Findings At Power, Exhibit 3, Barrier Integrity Screening Questions, dated June 19, 2012. The inspectors answered No to the Appendix A, Exhibit 3 barrier integrity screening questions, therefore, the finding was determined to be of very low safety significance (Green).

Licensee corrective actions included adding the RBFDS alarm function into the Maintenance Rule Program and performing an extent of condition review of all EOPs for SSC not scoped into the Maintenance Rule Program.

ATTACHMENT: