IR 05000346/2014002

IR 05000346-14-002; 1/1/14 - 3/31/14; Davis-Besse Nuclear Power Station; Fire Protection; Outage Activities

ML14113A073
Person / Time
Site:	Davis Besse
Issue date:	04/23/2014
From:	Jamnes Cameron Reactor Projects Region 3 Branch 4
To:	Lieb R FirstEnergy Nuclear Operating Co
References
IR-14-002
Download: ML14113A073 (55)

Text

UNITED STATES ril 23, 2014

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION NRC INTEGRATED INSPECTION REPORT 05000346/2014002

Dear Mr. Lieb:

On March 31, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Davis-Besse Nuclear Power Station. The enclosed report documents the results of this inspection, which were discussed with you and other members of your staff on April 15, 2014.

Based on the results of this inspection, two NRC-identified findings of very low safety significance were identified. Both of these findings also involved a violation of NRC requirements. However, because of their very low safety significance, and because the issues were entered into your corrective action program, the NRC is treating the issues as non-cited violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the subject or severity of any NCV, 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 NRC Resident Inspectors' Office at the Davis-Besse 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 Inspectors' Office at the Davis-Besse Nuclear Power Station. Additionally, as we informed you in the most recent NRC integrated inspection report, cross-cutting aspects identified in the last six months of 2013 using the previous terminology were being converted in accordance with the cross-reference in Inspection Manual Chapter (IMC) 0310. Section 4OA5.1 of the enclosed report documents the conversion of these cross-cutting aspects which will be evaluated for cross-cutting themes and potential substantive cross-cutting issues in accordance with IMC 0305 starting with the 2014 mid-cycle assessment review. If you disagree with the cross-cutting aspect assigned, 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 Inspectors' Office at the Davis-Besse 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)

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 No. 50-346 License No. NPF-3

Enclosure:

IR 05000346/2014002 w/Attachment: Supplemental Information

REGION III==

Docket No: 50-346 License No: NPF-3 Report No: 05000346/2014002 Licensee: FirstEnergy Nuclear Operating Company (FENOC)

Facility: Davis-Besse Nuclear Power Station Location: Oak Harbor, OH Dates: January 1 through March 31, 2014 Inspectors: D. Kimble, Senior Resident Inspector T. Briley, Resident Inspector J. Boettcher, Reactor Engineer M. Marshfield, Senior Resident Inspector - Perry Station M. Mitchell, Radiation Protection Inspector J. Neurauter, Senior Engineering Inspector J. Rutkowski, Project Engineer Approved by: J. Cameron, Chief Branch 4 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

Inspection Report (IR) 05000346/2014002; 1/1/14-3/31/14; Davis-Besse Nuclear Power Station;

Fire Protection; Outage Activities.

This report covers a three-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. Two Green findings were identified by the inspectors. Both findings were also considered non-cited violations (NCVs) 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, Components Within the Cross-Cutting Areas with an effective date of 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 4, dated December 2006.

NRC-Identified

and Self-Revealed Findings

Cornerstone: Initiating Events

Green.

The Inspectors identified a finding of very low safety significance and associated non-cited violations of Technical Specification (TS) 5.4.1(d) when the licensee failed to properly implement station procedures for control of ignition sources. Specifically, the inspectors identified two examples where the licensee did not adequately protect work areas containing combustible material from welding and grinding sparks generated in containment.

The finding was determined to be of more than minor significance because if left uncorrected would have the potential to lead to a more significant safety concern. In particular, uncontrolled ignition sources have the potential to start a fire that could impact risk significant plant equipment. The inspectors evaluated the finding using IMC 0609,

Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. Because the finding involved reactor shutdown operations and conditions, the inspectors transitioned to IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process - Phase 1 Operational Checklists for Both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). Since the finding was associated with an issue that occurred during the time the reactor was in a defueled condition, the inspectors conservatively consulted all four pressurized water reactor PWR checklists (i.e., Checklists 1 - 4). The inspectors determined that the finding did not adversely impact any shutdown defense-in-depth or mitigation attributes on any checklist, nor did it meet any of the checklist specific requirements for a Phase 2 or Phase 3 Significance Determination Process (SDP) analysis. Consequently, the finding was determined to be of very low safety significance. This finding had a cross-cutting aspect in the area of human performance associated with teamwork such that individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. In particular, the licensees standards and expectations for control of ignition sources were not adequately communicated to ensure adequate protection of combustible material. In the first example, the fire watch was unaware of the condition of the area below the welding activity. In the second example, the fire watch was generally unfamiliar with control of ignition source procedural requirements. In both cases, personnel passing by the work area observed hot sparks coming in contact with combustible material but did not communicate the condition to either the worker generating the sparks or the assigned fire watch to have the condition corrected. (H.4) (Section 1R05.1)

Cornerstone: Mitigating Systems

Green.

The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions,

Procedures, and Drawings, for the licensee's failure to maintain a containment "trash" gate closed and pinned while the area was unattended and the unit was in Mode 3.

Specifically, the inspectors identified Trash Gate 3, as referenced by plant procedure DB-OP-03013, Containment Daily Inspection and Containment Closeout Inspection, as being unpinned and open on February 1, 2014, when it should have been closed and pinned.

The finding was determined to be of more than minor significance because it was associated with the Mitigating Systems Cornerstone and directly impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to have the trash gate closed and pinned could allow debris generated during certain design basis accidents to degrade the capability of the Emergency Core Cooling System (ECCS) recirculation sump. The inspectors used Exhibit 2 - Mitigating Systems Screening Questions for mitigating systems, structures, components and functionality.

The finding screened out to be of very low safety significance because it was associated with a deficiency affecting the design or qualification of a mitigating system, structure, or component that did not result in a loss of operability or functionality. Specifically, the licensee had performed an analysis that concluded that the ECCS recirculation sump remained operable even with assuming additional debris reaching the upper sump screening in a post-accident environment. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution (PI&R) because the licensee's corrective actions for a previous issue were less than fully effective; the inspectors identified exactly the same issue under very similar circumstances in 2011 (see NCV 05000346/2011002-02 for additional details). (P.3) (Section 1R20.1)

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status

The unit began the inspection period operating at full power and, with the exception of small power maneuvers (e.g., reductions of 10 percent power or less) to facilitate planned evolutions and testing, remained operating at or near full power for the first month of the inspection period.

On February 1, 2014, the unit was shut down to begin its scheduled 18th refueling outage (see Section 1R20). On February 11, 2014, reactor defueling was completed to facilitate steam generator replacement activities, and the unit remained shut down with the reactor defueled for the balance of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness

1R01 Adverse Weather Protection

.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 Safety Analysis Report (USAR) 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 reviewed the impact of security modifications that could impede drainage from the site as discussed in recent licensee analysis of potential external flooding using current methodologies. The inspectors discussed aspects of this new flooding analysis with licensee personnel. Documents reviewed are listed in the to this report.

This review by the inspectors constituted a single external flooding sample as defined in Inspection Procedure (IP) 71111.01-05.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Quarterly Partial System Alignment Verifications

a. Inspection Scope

The inspectors performed partial system physical alignment verifications of the following risk-significant systems:

• the Motor Driven Feedwater Pump (MDFP) when auxiliary feedwater train 1 was out-of-service for a planned maintenance work window during the week ending January 18, 2014;

• Emergency Diesel Generator (EDG) No. 2 and the Station Blackout Diesel Generator (SBODG) during a period when EDG No. 1 was out-of-service for periodic testing and the 345 kilovolt (KV) Lemoyne transmission line was out-of-service for a planned maintenance work window during the week ending January 18, 2014; and

• decay heat (DH) train 1 when train 2 was out-of-service for planned maintenance work window during the weeks ending March 8, 2014 and March 15, 2014.

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, USAR, Technical Specification (TS) requirements, outstanding work orders (WOs), condition reports (CRs), 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 corrective action program (CAP) with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.

These activities by the inspectors constituted three partial system alignment verification inspection samples as defined in IP 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Routine Resident Inspector Tours

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:

• clean waste receiver tank room No. 1 and No. 2 (Auxiliary Building, rooms 123 and 124);

• steam generator west D-ring area (Containment, room 216);

• steam generator east D-ring area (Containment, room 218);

• 653 elevation (Containment, rooms 700 and 701);

• 603 elevation - containment vessel heating pads (Containment, room 407); and

• the Shield Building construction opening external work platform.

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.

In addition to standard refueling outage (RFO) activities, activities during the site's 18th RFO also included replacement of the unit's two steam generators and significant portions of the connected reactor coolant system (RCS) piping. Inspection of the steam generator replacement and associated activities is covered under NRC IP 50001, "Steam Generator Replacement Inspection," and will be documented in a separate NRC Inspection Report 05000346/2013010. Applicable portions of the inspectors' reviews documented in this section also were credited toward completion of IP 50001.

Documents reviewed are listed in the Attachment to this report.

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

b. Findings

Failure to Implement Fire Protection Plan Requirements Related to Control of Ignition Sources Introduction An NRC-identified finding of very low safety significance (Green) and associated NCV of TS 5.4.1(d) were identified when the licensee failed to properly implement station procedures for control of ignition sources. Specifically, the inspectors identified two examples where the licensee did not adequately protect work areas containing combustible material from welding and grinding sparks generated in containment.

Description On February 19, 2014, the inspectors observed weld sparks in containment at the 565 elevation coming down from the overhead. Upon further investigation, licensee contract workers were welding lifting lugs onto a section of feedwater piping on the 610 elevation above. An assigned fire watch was present observing the welding activity; however, neither the fire watch nor the welder was aware of the weld sparks falling to the work area below. The sparks were observed to be coming in contact with other contract workers wearing non-fire retardant anti-contamination clothing along with some plastic bags stored on the 565 elevation. The supervisor for the welding activity was notified of the condition; however, the welding activity had already completed prior to their arrival at the work area. CR 2014-03305 was initiated to document the issue.

On March 1, 2014, the inspectors observed sparks from grinding a metal surface (section of upper lateral restraint for steam generator No. 1) going over the flame-resistant safety barrier being used to protect combustible material from flying sparks. The sparks had come in contact with other workers wearing non-fire retardant anti-contamination clothing as they passed by the work area at the 653 elevation in containment, in addition to other miscellaneous combustible materials (floor matting, tape, rope, etc.). An assigned fire watch was present monitoring the grinding activity.

When questioned by the inspectors, the fire watch was cognizant of the sparks coming over the flame-resistant safety barrier and indicated that any fires that would start as a result of the condition would be put out right away. The fire watch was unaware of the procedural requirements for control of ignition sources such that any combustible material within a 35 foot radius or below the work area would be protected with flame-resistant material to prevent a fire from starting. A licensee safety representative was informed of the condition, immediately stopped the grinding activity, and repositioned the flame-resistant safety barrier prior to work resuming. CR 2014-04128 was initiated to document the issue.

Analysis The inspectors reviewed this finding using the guidance contained in Appendix B, Issue Screening, of Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports. The inspectors determined that the licensees failure to properly implement plant procedures for controlling ignition sources was a performance deficiency that was reasonably within the licensees ability to foresee and correct and should have been prevented. This finding was associated with the Initiating Events Cornerstone of reactor safety and was of more than minor significance because if left uncorrected would have the potential to lead to a more significant safety concern. In particular, uncontrolled ignition sources have the potential to start a fire that could impact risk significant plant equipment.

The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings. Because the finding involved reactor shutdown operations and conditions, the inspectors transitioned to IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process - Phase 1 Operational Checklists for Both PWRs and BWRs. Since the finding was associated with an issue that occurred during the time the reactor was in a defueled condition, the inspectors conservatively consulted all four PWR checklists (i.e., Checklists 1 - 4). The inspectors determined that the finding did not adversely impact any shutdown defense-in-depth or mitigation attributes on any checklist, nor did it meet any of the checklist specific requirements for a Phase 2 or Phase 3 SDP analysis. Consequently, the finding was determined to be of very low safety significance (Green).

This finding had a cross-cutting aspect in the area of human performance associated with teamwork such that individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. In particular, the licensees standards and expectations for control of ignition sources were not adequately communicated to ensure adequate protection of combustible material. In the first example, the fire watch was unaware of the condition of the area below the welding activity. In the second example, the fire watch was generally unfamiliar with control of ignition source procedural requirements. In both cases, personnel passing by the work area observed hot sparks coming in contact with combustible material but did not communicate the condition to either the worker generating the sparks or the assigned fire watch to have the condition corrected. (H.4)

Enforcement TS 5.4.1(d), requires, in part, the licensee to establish, implement, and maintain applicable written procedures covering fire protection program implementation. The fire protection program was implemented, in part, by Davis-Besse procedure DB-FP-00018, Control of Ignition Sources, Revision 12. Procedure DB-FP-00018, step 6.1.5, states that all movable combustible material below and within a 35 foot horizontal radius of the work area shall be removed or protected with flame resistant material; for example, fire retardant paint on wood scaffolding, fire resistant blankets covering cables, etc. Step 6.1.7 states Welding blankets shall be used to cover any exposed cables, flammable liquid, paper, rags, other combustible materials, or open paths between rooms within the 35 ft. radius or below the work area where the possibility exists of exposure to flame, sparks or slag. Contrary to DB-FP-00018, on February 19, 2014 and March 1, 2014, the licensee failed to control ignition sources such that all movable and combustible material below and within a 35 foot horizontal radius of the work area were removed or covered with welding blankets. Because this finding is of very low safety significance (Green), had been entered into the licensees CAP, and the licensee had taken or planned corrective actions under CRs 2014-03305 and 2014-04128, the associated violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. Corrective actions taken by the licensee include, but are not limited to, a stand-down with contract fire watch personnel to reinforce expectations and review requirements for control of ignition sources, and communication of lessons learned during work group turnovers. (NCV 05000346/2014002-01)

1R11 Licensed Operator Requalification Program

.1 Resident Inspector Quarterly Review of Licensed Operator Simulator Training

a. Inspection Scope

On January 8, 2014, the inspectors observed an integrated crew of licensed and non-licensed operators in the plants simulator during a training simulator scenario. The training scenario observed was part of the licensee's pre-outage licensed and non-licensed operator training prior to the 18th RFO, and involved multiple complex cold shutdown casualties with a serious, large area station fire.

The inspectors verified that operator performance was adequate, that evaluators were identifying and documenting crew performance problems, and that training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• licensed operator performance;

• the clarity and formality of communications;

• the ability of the crew to take timely and conservative actions;

• the crews prioritization, interpretation, and verification of annunciator alarms;

• the correct use and implementation of abnormal and emergency procedures by the crew;

• control board manipulations;

• the oversight and direction provided by licensed Senior Reactor Operators (SROs); and

• the ability of the crew to identify and implement appropriate TS actions and Emergency Plan (EP) actions and notifications.

The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements.

In addition to standard RFO activities, activities during the site's 18th RFO also included replacement of the unit's two steam generators and significant portions of the connected RCS piping. Inspection of the steam generator replacement and associated activities is covered under NRC IP 50001, "Steam Generator Replacement Inspection," and will be documented in a separate NRC Inspection Report 05000346/2013010. Applicable portions of the inspectors' reviews documented in this section also were credited toward completion of IP 50001. Documents reviewed are listed in the Attachment to this report.

These observations and activities by the inspectors constituted a single quarterly licensed operator requalification program simulator training inspection sample as defined in IP 71111.11-05.

b. Findings

No findings were identified.

.2 Resident Inspector Quarterly Observation of Control Room Activities

a. Inspection Scope

During the course of the inspection period, the inspectors performed numerous observations of licensed operator performance in the plants control room to verify that operator performance was adequate and that plant evolutions were being conducted in accordance with approved plant procedures. Specific activities observed that involved a heightened tempo of activities or periods of elevated risk included, but were not limited to:

• end-of-cycle reductions in RCS average temperature (Tave) to maximize nuclear fuel burn-up prior to the plant's 18th RFO during the weeks ending January 11 ,2014, through January 25, 2014;

• circulating pump and cooling tower realignment for extreme cold weather operations during the week ending January 25, 2014;

• reactor shutdown for plant's 18th RFO during the week ending February 1, 2014;

• reactor cool down and RCS depressurization, including the blocking of safety features actuation system (SFAS) low pressure trips and isolation of core flood tanks, during the week ending February 8, 2014;

• initiation of DH removal cooling on the RCS and the securing of reactor coolant pumps during the week ending February 8, 2014; and

• selected portions of integrated SFAS response time testing during the week ending March 1, 2014.

The inspectors evaluated the following areas during the course of the control room observations:

• licensed operator performance;

• the clarity and formality of communications;

• the ability of the crew to take timely and conservative actions;

• the crews prioritization, interpretation, and verification of annunciator alarms;

• the correct use and implementation of normal operating, annunciator alarm response, and abnormal operating procedures by the crew;

• control board manipulations;

• the oversight and direction provided by on-watch SROs and plant management personnel; and

• the ability of the crew to identify and implement appropriate TS actions and notifications.

The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements.

In addition to standard RFO activities, activities during the site's 18th RFO also included replacement of the unit's two steam generators and significant portions of the connected RCS piping. Inspection of the steam generator replacement and associated activities is covered under NRC IP 50001, "Steam Generator Replacement Inspection," and will be documented in a separate NRC Inspection Report 05000346/2013010. Applicable portions of the inspectors' reviews documented in this section also were credited toward completion of IP 50001. Documents reviewed are listed in the Attachment to this report.

These observation activities by the inspectors of operator performance in the stations control room constituted a single quarterly inspection sample as defined in IP 71111.11-05.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

.1 Routine Quarterly Evaluations

a. Inspection Scope

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

• heat tracing systems and features; and

• core flood tanks and associated equipment.

The inspectors reviewed events such as where ineffective equipment maintenance could result in or had resulted in valid or invalid automatic actuations or system transients 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 systems, structures, 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.

The inspectors reviews constituted two quarterly maintenance effectiveness inspection samples as defined in IP 71111.12-05.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

.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:

• planned work associated with rigging and lifting the reactor vessel closure head during the week ending February 8, 2014;

• licensee response to the inadvertent isolation of a DH system suction line pressure gauge being used for RCS draining during the week ending February 8, 2014;

• licensee response to a concrete void identified in the shield building concrete during the week ending February 15, 2014;

• planned work associated with rigging and heavy load lift for steam generator No. 1-1 during the week ending March 1, 2014; and

• planned work associated with rigging and heavy load lift for steam generator No. 1-2 during the weeks ending March 8, 2014, and March 15, 2014.

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.

In addition to standard RFO activities, activities during the site's 18th RFO also included replacement of the unit's two steam generators and significant portions of the connected RCS piping. Inspection of the steam generator replacement and associated activities is covered under NRC IP 50001, "Steam Generator Replacement Inspection," and will be documented in a separate NRC Inspection Report 05000346/2013010. Applicable portions of the inspectors' reviews documented in this section also were credited toward completion of IP 50001. Documents reviewed are listed in the Attachment to this report.

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

b. Findings

No findings were identified.

1R15 Operability Determinations and Functional Assessments

.1 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• operability of the high pressure injection (HPI) recirculation line to the borated water storage tank (BWST) during subzero temperatures, as documented in CR 2014-00190, during the week ending January 18, 2014;

• operability of an HPI Train with an inoperable or isolated DH heat exchanger, as documented in CR 2013-18634, during the week ending February 1, 2014;

• operability evaluation associated with reactor vessel closure head temperature effects on plant transients for Cycle 19, as documented in CR 2014-00549, during the weeks ending January 18, 2014, and January 25, 2014;

• operability evaluation associated with a concrete void identified in the shield building concrete, as documented in CR 2014-02896, during the weeks ending February 15, 2014, through March 29, 2014; and

• operability evaluation associated with several broken steel rebar segments identified during shield building hydrodemolition activities, as documented in CR 2014-02482, during the weeks ending February 15, 2014, through March 29, 2014.

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 USAR 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.

In addition to standard RFO activities, activities during the site's 18th RFO also included replacement of the unit's two steam generators and significant portions of the connected RCS piping. Inspection of the steam generator replacement and associated activities is covered under NRC IP 50001, "Steam Generator Replacement Inspection," and will be documented in a separate NRC Inspection Report 05000346/2013010. Applicable portions of the inspectors' reviews documented in this section also were credited toward completion of IP 50001. Documents reviewed are listed in the Attachment to this report.

The review of these issues by the inspectors constituted five inspection samples as defined in IP 71111.15-05.

b. Findings

No findings were identified.

1R18 Plant Modifications

.1 Temporary Plant Modification

a. Inspection Scope

The inspectors reviewed the following temporary modification to the facility:

• Engineering Change Package (ECP) No. 12-0421: License Renewal Commitment - Core Bore in Concrete Floor at Elevation 537' Inside Containment.

The inspectors reviewed the configuration changes and associated 10 CFR 50.59 safety evaluation documents against the design basis, the USAR, and the TS, as applicable, to verify that the modification did not affect the operability or availability of any safety-related systems, or systems important to safety. The inspectors observed ongoing and completed work activities to ensure that the modification was installed as directed and consistent with the design control documents; that the modification operated as expected; and that operation of the modification did not impact the operability of any interfacing systems. The inspectors verified that relevant procedure, design, and licensing documents were properly updated both for the interim condition and after the temporary changes to the facility were restored to the original configuration.

Lastly, the inspectors discussed the plant modification with Operations and Engineering personnel to ensure that the individuals were aware of how the operation with the modification in place could impact overall plant performance. Documents reviewed in the course of this inspection are listed in the Attachment to this report.

The inspectors review of this temporary plant modification constituted a single inspection sample as defined in IP 71111.18-05.

b. Findings

No findings were identified.

.2 Permanent Plant Modification

a. Inspection Scope

The inspectors reviewed the following modification:

• ECP No. 12-0695: Replacement of the Control Room main access door (Door No. 509).

The inspectors reviewed the configuration changes and associated 10 CFR 50.59 safety evaluation documents against the design basis, the USAR, and the TS, as applicable, to verify that the modification did not affect the operability or availability of any safety-related systems, or systems important to safety. The inspectors observed ongoing and completed work activities to ensure that the modification was installed as directed and consistent with the design control documents; that the modification operated as expected; and that operation of the modification did not impact the operability of any interfacing systems. The inspectors verified that relevant procedure, design, and licensing documents were properly updated. Finally, the inspectors discussed the plant modification with Operations, Engineering, Security, and Training Department personnel to ensure that the individuals were aware of how the operation with the modification in place could impact overall plant performance. Documents reviewed in the course of this inspection are listed in the Attachment to this report.

The inspectors review of this permanent plant modification constituted a single inspection sample as defined in IP 71111.18-05.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

.1 Quarterly Resident Inspector Observation and Review of Post-Maintenance Testing

Activities

a. Inspection Scope

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

• operational testing of auxiliary feed pump No.1 following planned preventative maintenance (PM) to replace Agastat relays during the week ending January 18, 2014;

• functional testing of the H14 containment auxiliary crane (Palfinger) following emergent replacement with a temporary crane unit during the week ending February 22, 2014;

• functional testing of EDG No. 1 following planned replacement of the engine lube oil cooler during the week ending March 8, 2014;

• operational testing of DH Pump No. 2 following planned replacement of the inboard pump seal during the week ending March 15, 2014;

• functional testing of the control room envelope following planned replacement of control room door 509 during the week ending March 22, 2014;

• operational testing of relays and lockout functions for startup transformer X01 following extensive modifications to the plant switchyard during the week ending March 29, 2014; and

• functional testing of the ability of the Ohio Edison 345 KV power line to manually and automatically support side loads following extensive modifications to the plant switchyard during the week ending March 29, 2014.

These activities were selected based upon the system, structure 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 USAR, 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 the PMTs 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.

In addition to standard RFO activities, activities during the site's 18th RFO also included replacement of the unit's two steam generators and significant portions of the connected RCS piping. Inspection of the steam generator replacement and associated activities is covered under NRC IP 50001, "Steam Generator Replacement Inspection," and will be documented in a separate NRC Inspection Report 05000346/2013010. Applicable portions of the inspectors' reviews documented in this section also were credited toward completion of IP 50001. Documents reviewed are listed in the Attachment to this report.

The inspectors reviews of these activities constituted seven PMT inspection samples as defined in IP 71111.19-05.

b. Findings

No findings were identified.

1R20 Outage Activities

.1 Refueling Outage Activities

a. Inspection Scope

The inspectors reviewed the licensee's comprehensive outage plan, shutdown defense-in-depth plan, and contingencies for the plant's 18th RFO, which began on February 1, 2014, and continued through the end of the inspection period. These reviews were performed to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. During the RFO, the inspectors observed portions of the shutdown and RCS cool down and depressurization, and monitored licensee controls over the outage activities listed below:

• licensee configuration management, including maintenance of defense-in-depth commensurate with the shutdown defense-in-depth plan for key safety functions and compliance with the applicable TS when taking equipment out of service;

• implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing;

• installation and configuration of RCS pressure, level, and temperature instruments to provide accurate indication, accounting for instrument error;

• controls over the status and configuration of electrical systems to ensure that TS and shutdown defense-in-depth plan requirements were met, and controls over switchyard activities;

• monitoring of DH removal processes, systems, and components;

• controls to ensure that outage work was not impacting the ability of the operators to operate the spent fuel pool cooling system;

• reactor water inventory controls including flow paths, configurations, and alternative means for inventory addition, and controls to prevent inventory loss;

• controls over activities that could affect reactivity;

• maintenance of containment and associated ventilation systems, as required by TS;

• licensee fatigue management, as required by 10 CFR 26, Subpart I;

• refueling activities, including fuel handling, spent fuel assembly inspections, and fuel assembly reconstitution; and

• licensee identification and resolution of problems related to RFO activities.

In addition to standard RFO activities, activities during the site's 18th RFO also included replacement of the unit's two steam generators and significant portions of the connected RCS piping. Inspection of the steam generator replacement and associated activities is covered under NRC IP 50001, "Steam Generator Replacement Inspection," and will be documented in a separate NRC Inspection Report 05000346/2013010. Applicable portions of the inspectors' reviews documented in this section also were credited toward completion of IP 50001. Documents reviewed are listed in the Attachment to this report.

Because the RFO was still ongoing at the end of the inspection period, these RFO review activities constituted only a partial RFO inspection sample as defined in IP 71111.20-05.

b. Findings

Containment Emergency Core Cooling System Recirculation Sump Reliability Degraded Due to Unfastened Debris Gate Introduction A finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensee's failure to maintain a containment "trash" gate closed and pinned while the area was unattended and the unit was in Mode 3.

Description The licensee has several "trash" gates in the containment that were designed to help minimize post-accident debris loading on the emergency core cooling system (ECCS)recirculation sump inlet screens. Several of these gates are located on the 565' elevation of containment, which is the same elevation as the top of the containment post-accident ECCS recirculation sump. The gates support licensee analyses for steam line break and loss of coolant accident scenarios. In these scenarios debris from containment coatings and piping insulation can be generated and ultimately washed down into the post-accident ECCS recirculation sump, challenging ECCS pump operability.

In Modes 1 through 4, when the ECCS recirculation sump is required to be operable, the licensee specified that the gates shall be closed, and if so equipped, pinned in the closed position except for passage and if personnel are working inside the gates.

Procedure DB-OP-03013, Containment Daily Inspection and Containment Closeout Inspection, contained directions to have Trash Gates 1 and 3 closed, with load pins installed and held in place by cotter pins.

In the early afternoon on February 1, 2014, with the unit shut down and in Mode 3, the inspectors conducted a visual inspection of the general condition of the containment interior. While inspecting conditions on the containment 565' elevation, the inspectors found trash gate 3 unpinned and open. A search of the area by the inspectors revealed no personnel in the vicinity. Prior to leaving the area, the inspectors closed and pinned trash gate 3, as specified by the DB-OP-03013 requirements.

Typically, the licensee will send a number of personnel into containment shortly after reactor shutdown on the first day of a RFO. Personnel and tasks include radiological surveys by radiation protection (RP) technicians, craft workers to rig lights and build scaffolding, engineering personnel conducting visual inspections, operations personnel conducting system and equipment alignments, etc. For the issue with trash gate 3 being found unpinned and open on February 1, 2014, the licensee determined that it was most probably the result of workers building scaffolding in the area a short time before the inspectors had identified the issue and that they had forgotten to secure the gate upon exiting the work area.

Analysis The inspectors determined that the failure of licensee personnel to close and pin Trash Gate 3 after working in the area on February 1, 2014, was contrary to established plant procedural requirements, and constituted a performance deficiency that was reasonably within the licensee's ability to foresee and correct and which should have been prevented.

The finding was determined to be of more than minor significance because it was associated with the Mitigating Systems Cornerstone of Reactor Safety and directly impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, failure to have the trash gate closed and pinned could allow debris generated during certain design basis accidents to degrade the capability of the ECCS recirculation sump.

The inspectors determined the finding could be evaluated using the Significance SDP in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors used Exhibit 2 - Mitigating Systems Screening Questions for mitigating systems, structures, components and functionality. The finding screened out to be of very low safety significance (Green) because it was associated with a deficiency affecting the design or qualification of a mitigating system, structure, or component that did not result in a loss of operability or functionality. Specifically, the licensee had performed an analysis that concluded that the ECCS recirculation sump remained operable even with assuming additional debris reaching the upper sump screening in a post-accident environment.

This finding has a cross-cutting aspect in the area of PI&R because the licensee's corrective actions for a previous issue were less than fully effective. Specifically, the inspectors identified exactly the same issue under very similar circumstances in 2011 (see NCV 05000346/2011002-02 for additional details). (P.3)

Enforcement 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Contrary to this requirement, on February 1, 2014, the licensee failed to comply with a requirement in a station procedure. Specifically, while the plant was in Mode 3, licensee personnel failed to close and pin trash gate 3 in containment after completing a work activity in the area and exiting through that gate. Because this violation was of very low safety significance and had been entered into the licensees CAP as CR 2014-01736, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000346/2014002-02)

1R22 Surveillance Testing

.1 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the results for the following testing 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:

• initial receipt inspections/tests associated with the arrival of new fuel assemblies during the week ending January 11, 2014 (routine);

• emergency diesel generator No. 1 normal periodic monthly TS function and operability testing during the week ending January 18, 2014 (routine);

• hydrogen seal oil room fire detector and deluge testing during the week ending January 25, 2014 (routine);

• main steam safety valve setpoint testing during the week ending February 1, 2014 (inservice testing (IST));

• containment equipment hatch emergency closure testing during the week ending February 8, 2014 (routine);

• local leak rate testing of penetration 16, containment vent header isolation valve (RC1719A), during the weeks ending February 15, 2014 and February 22, 2014 (containment isolation valve (CIV)); and

• service water pump 3 shutdown testing during the week ending March 22, 2014 (IST).

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;

• that measuring and test equipment calibration was current;

• that test equipment was used within the required range and accuracy;

• that applicable prerequisites described in the test procedures were satisfied;

• that 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;

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

• that test equipment was removed after testing;

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

• where applicable for safety-related instrument control surveillance tests, that reference setting data were accurately incorporated in the test procedure;

• where applicable, that actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished;

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

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

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

Documents reviewed are listed in the Attachment to this report.

These activities conducted by the inspectors constituted four routine surveillance testing inspection samples, two IST inspection samples, and one CIV inspection sample as defined in IP 71111.22, Sections -02 and -05.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Occupational Radiation Safety and Public Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

The inspectors' activities documented in this section constituted a partial inspection sample as defined in IP 71124.01-05.

.1 Inspection Planning (02.01)

a. Inspection Scope

The inspectors reviewed all licensee performance indicators (PIs) for the Occupational Radiation Safety Cornerstone for follow-up. The inspectors reviewed the results of RP Program audits (e.g., licensees quality assurance audits or other independent audits).

The inspectors reviewed any reports of operational occurrences related to occupational radiation safety since the last inspection. The inspectors reviewed the results of the audit and operational report reviews to gain insights into overall licensee performance.

b. Findings

No findings were identified.

.2 Radiological Hazard Assessment (02.02)

a. Inspection Scope

The inspectors determined if there had been changes to plant operations since the last inspection that may result in a significant new radiological hazard for onsite workers or members of the public. The inspectors evaluated whether the licensee assessed the potential impact of these changes and has implemented periodic monitoring, as appropriate, to detect and quantify the radiological hazard.

The inspectors reviewed the last two radiological surveys from selected plant areas and evaluated whether the thoroughness and frequency of the surveys where appropriate for the given radiological hazard.

The inspectors conducted walkdowns of the facility; including radioactive waste processing, storage, and handling areas to evaluate material conditions and performed independent radiation measurements to verify conditions.

The inspectors selected the following radiologically risk-significant work activities that involved exposure to radiation:

• containment scaffold building;

• reactor canal leak mitigation;

• containment structural interferences; and

• RCS cutting and preparing.

For these work activities, the inspectors assessed whether the pre-work surveys performed were appropriate to identify and quantify the radiological hazard and to establish adequate protective measures. The inspectors evaluated the Radiological Survey Program to determine if hazards were properly identified, including the following:

• identification of hot particles;

• the presence of alpha emitters;

• the potential for airborne radioactive materials, including the potential presence of transuranics and/or other hard-to-detect radioactive materials (this evaluation may include licensee-planned entry into non-routinely entered areas subject to previous contamination from failed fuel.);

• the hazards associated with work activities that could suddenly and severely increase radiological conditions and that the licensee has established a means to inform workers of changes that could significantly impact their occupational dose; and

• severe radiation field dose gradients that can result in non-uniform exposures of the body.

The inspectors observed work in potential airborne areas and evaluated whether the air samples were representative of the breathing air zone. The inspectors evaluated whether continuous air monitors were located in areas with low background to minimize false alarms and were representative of actual work areas. The inspectors evaluated the licensees program for monitoring levels of loose surface contamination in areas of the plant with the potential for the contamination to become airborne.

b. Findings

No findings were identified.

.3 Instructions to Workers (02.03)

a. Inspection Scope

The inspectors selected various containers holding non-exempt licensed radioactive materials that may cause unplanned or inadvertent exposure of workers, and assessed whether the containers were labeled and controlled in accordance with 10 CFR 20.1904, Labeling Containers, or met the requirements of 10 CFR 20.1905(g), Exemptions To Labeling Requirements.

The inspectors reviewed the following radiation work permits (RWPs) used to access high radiation areas and evaluated the specified work control instructions or control barriers:

• RWP 2014-5205; Containment Scaffold Building; Revision 0;

• RWP 2014-5110; Reactor Canal Leak Mitigation; Revision 0;

• RWP 2014-5217; Containment Structural Interferences; Revision 0; and

• RWP 2014-5212; Reactor Coolant System Cutting and Preparing; Revision 0.

For these RWPs, the inspectors assessed whether allowable stay times or permissible dose (including from the intake of radioactive material) for radiologically significant work under each RWP were clearly identified. The inspectors evaluated whether electronic personal dosimeter alarm set-points were in conformance with survey indications and plant policy.

For work activities that could suddenly and severely increase radiological conditions, the inspectors assessed the licensees means to inform workers of changes that could significantly impact their occupational dose.

b. Findings

No findings were identified.

.4 Contamination and Radioactive Material Control (02.04)

a. Inspection Scope

The inspectors observed locations where the licensee monitors potentially contaminated material leaving the radiological control area and inspected the methods used for control, survey, and release from these areas. The inspectors observed the performance of personnel surveying and releasing material for unrestricted use and evaluated whether the work was performed in accordance with plant procedures and whether the procedures were sufficient to control the spread of contamination and prevent unintended release of radioactive materials from the site. The inspectors assessed whether the radiation monitoring instrumentation had appropriate sensitivity for the types of radiation present.

The inspectors reviewed the licensees criteria for the survey and release of potentially contaminated material. The inspectors evaluated whether there was guidance on how to respond to an alarm that indicates the presence of licensed radioactive material.

The inspectors reviewed the licensees procedures and records to verify that the radiation detection instrumentation was used at its typical sensitivity level based on appropriate counting parameters. The inspectors assessed whether or not the licensee had established a de facto release limit by altering the instruments typical sensitivity through such methods as raising the energy discriminator level or locating the instrument in a high radiation background area.

b. Findings

No findings were identified.

.5 Radiological Hazards Control and Work Coverage (02.05)

a. Inspection Scope

The inspectors evaluated ambient radiological conditions (e.g., radiation levels or potential radiation levels) during tours of the facility. The inspectors assessed whether the conditions were consistent with applicable posted surveys, RWPs, and worker briefings.

The inspectors evaluated the adequacy of radiological controls, such as required surveys, RP job coverage (including audio and visual surveillance for remote job coverage), and contamination controls. The inspectors evaluated the licensees use of electronic personal dosimeters in high noise areas as high radiation area monitoring devices.

The inspectors assessed whether radiation monitoring devices were placed on the individuals body consistent with the licensees procedures. The inspectors assessed whether the dosimeter was placed in the location of highest expected dose or that the licensee properly employed an NRC-approved method of determining effective dose equivalent.

The inspectors reviewed the application of dosimetry to effectively monitor exposure to personnel in high radiation work areas with significant dose rate gradients.

The inspectors checked for RWPs for work within airborne radioactivity areas with the potential for individual worker internal exposures. There were no airborne radioactivity RWPs planned and issued; all work was less than 0.3 Derived Air Concentration.

b. Findings

No findings were identified.

.6 Risk-Significant High Radiation Area and Very High Radiation Area Controls (02.06)

a. Inspection Scope

The inspectors evaluated licensee controls for very high radiation areas and areas with the potential to become very high radiation areas to ensure that an individual was not able to gain unauthorized access to the very high radiation areas.

b. Findings

No findings were identified.

.7 Radiation Worker Performance (02.07)

a. Inspection Scope

The inspectors observed radiation worker performance with respect to stated RP work requirements. The inspectors assessed whether workers were aware of the radiological conditions in their workplace and the RWP controls/limits in place, and whether their performance reflected the level of radiological hazards present.

b. Findings

No findings were identified.

.8 Radiation Protection Technician Proficiency (02.08)

a. Inspection Scope

The inspectors observed the performance of the RP technicians with respect to all RP work requirements. The inspectors evaluated whether technicians were aware of the radiological conditions in their workplace and the RWP controls/limits, and whether their performance was consistent with their training and qualifications with respect to the radiological hazards and work activities.

b. Findings

No findings were identified.

2RS2 Occupational As-Low-As-Reasonably-Achievable Planning and Controls

The inspectors' activities documented in this section constituted a partial inspection sample as defined in IP 71124.02-05.

.1 Inspection Planning (02.01)

a. Inspection Scope

The inspectors reviewed pertinent information regarding plant collective exposure history, current exposure trends, and ongoing or planned activities in order to assess current performance and exposure challenges. The inspectors reviewed the plants three-year rolling average collective exposure.

The inspectors reviewed the site-specific trends in collective exposures and source term measurements.

The inspectors reviewed site-specific procedures associated with maintaining occupational exposures as-low-as-reasonably-achievable (ALARA), which included a review of processes used to estimate and track exposures from specific work activities.

b. Findings

No findings were identified.

.2 Radiological Work Planning (02.02)

a. Inspection Scope

The inspectors selected the following work activities of the highest exposure significance:

• RWP 2014-5205; Containment Scaffold Building; Revision 0;

• RWP 2014-5110; Reactor Canal Leak Mitigation; Revision 0;

• RWP 2014-5217; Containment Structural Interferences; Revision 0; and

• RWP 2014-5212; Reactor Coolant System Cutting and Preparing; Revision 0.

The inspectors reviewed the ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements. The inspectors determined whether the licensee reasonably grouped the radiological work into work activities based on historical precedence, industry norms, and/or special circumstances.

The inspectors assessed whether the licensees planning identified appropriate dose mitigation features, considered alternate mitigation features, and defined reasonable dose goals. The inspectors evaluated whether the licensees ALARA assessment has taken into account decreased worker efficiency from use of respiratory protective devices and/or heat stress mitigation equipment (e.g., ice vests). The inspectors determined whether the licensees work planning considered the use of remote technologies (e.g., teledosimetry, remote visual monitoring, and robotics) as a means to reduce dose and the use of dose reduction insights from industry operating experience and plant-specific lessons learned. The inspectors assessed the integration of ALARA requirements into work procedure and RWP documents.

b. Findings

No findings were identified.

.3 Verification of Dose Estimates and Exposure Tracking Systems (02.03)

a. Inspection Scope

The inspectors reviewed the assumptions and basis (including dose rate and man-hour estimates) for the current annual collective exposure estimate for reasonable accuracy for select ALARA work packages. The inspectors reviewed applicable procedures to determine the methodology for estimating exposures from specific work activities and the intended dose outcome.

The inspectors evaluated whether the licensee established measures to track, trend, and, if necessary, to reduce occupational doses for ongoing work activities. The inspectors assessed whether trigger points or criteria were established to prompt additional reviews and/or additional ALARA planning and controls.

The inspectors evaluated the licensees method of adjusting exposure estimates, or re-planning work, when unexpected changes in scope or emergent work were encountered. The inspectors assessed whether adjustments to exposure estimates (i.e., intended dose) were based on sound RP and ALARA principles or if they were just adjusted to account for failures to control the work. The inspectors evaluated whether the frequency of these adjustments called into question the adequacy of the original ALARA planning process.

b. Findings

No findings were identified.

.4 Radiation Worker Performance (02.05)

a. Inspection Scope

The inspectors observed radiation worker and RP technician performance during work activities being performed in radiation areas, airborne radioactivity areas, or high radiation areas. The inspectors evaluated whether workers demonstrated the ALARA philosophy in practice (e.g., workers are familiar with the work activity scope and tools to be used and workers used ALARA low-dose waiting areas) and whether there were any procedure compliance issues (e.g., workers are not complying with work activity controls). The inspectors observed radiation worker performance to assess whether the training and skill level was sufficient with respect to the radiological hazards and the work involved.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

The inspectors' activities documented in this section constituted a partial inspection sample as defined in IP 71124.03-05.

.1 Engineering Controls (02.02)

a. Inspection Scope

The inspectors reviewed the licensees use of permanent and temporary ventilation to determine whether the licensee uses ventilation systems as part of its engineering controls (in lieu of respiratory protection devices) to control airborne radioactivity. The inspectors reviewed procedural guidance for use of installed plant systems, such as containment purge, spent fuel pool ventilation, and auxiliary building ventilation, and assessed whether the systems are used, to the extent practicable, during high-risk activities (e.g., using containment purge during cavity floodup).

The inspectors selected installed ventilation systems used to mitigate the potential for airborne radioactivity and evaluated whether the ventilation airflow capacity, flow path (including the alignment of the suction and discharges), and filter/charcoal unit efficiencies, as appropriate, were consistent with maintaining concentrations of airborne radioactivity in work areas below the concentrations of an airborne area to the extent practicable.

The inspectors selected temporary ventilation system setups (high-efficiency particulate air/charcoal negative pressure units, down draft tables, tents, metal Kelly buildings, and other enclosures) used to support work in contaminated areas. The inspectors assessed whether the use of these systems is consistent with licensee procedural guidance and ALARA concept.

The inspectors reviewed airborne monitoring protocols by selecting installed systems used to monitor and warn of changing airborne concentrations in the plant and evaluated whether the alarms and setpoints were sufficient to prompt licensee/worker action to ensure that doses are maintained within the limits of 10 CFR Part 20 and the ALARA concept.

The inspectors assessed whether the licensee established trigger points (e.g., the Electric Power Research Institutes Alpha Monitoring Guidelines for Operating Nuclear Power Stations) for evaluating levels of airborne beta-emitting (e.g., plutonium-241) and alpha-emitting radionuclides.

b. Findings

No findings were identified.

2RS8 Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and

Transportation (71124.08)

The inspectors' activities documented in this section constituted a partial inspection sample as defined in IP 71124.08-05.

.1 Radioactive Material Storage (02.02)

a. Inspection Scope

The inspectors assessed whether the radioactive material storage area for the old steam generators were controlled and posted in accordance with the requirements of 10 CFR Part 20, Standards for Protection against Radiation. For materials stored or used in the controlled or unrestricted areas, the inspectors evaluated whether they were secured against unauthorized removal and controlled in accordance with 10 CFR 20.1801, Security of Stored Material, and 10 CFR 20.1802, Control of Material Not in Storage, as appropriate.

The inspectors evaluated whether the licensee established a process for monitoring the impact of long-term storage (e.g., buildup of any gases produced by waste decomposition, chemical reactions, container deformation, loss of container integrity, or re-release of free-flowing water) that was sufficient to identify potential unmonitored, unplanned releases or nonconformance with waste disposal requirements.

b. Findings

No findings were identified.

OTHER ACTIVITIES

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

4OA1 Performance Indicator Verification

.1 Unplanned Scrams per 7000 Critical Hours

a. Inspection Scope

The inspectors sampled licensee submittals for the Unplanned Scrams per 7000 Critical Hours PI for the period from January 2013 to December 2013. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the Nuclear Energy Institute (NEI) Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, were used. The inspectors reviewed the licensees operations narrative logs, CRs, event reports and NRC integrated IRs for the period to validate the accuracy of the submittals.

The inspectors also reviewed the licensees CAP to determine if any problems had been identified with the PI data collected or transmitted for this indicator. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one Unplanned Scrams per 7000 Critical Hours sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.2 Unplanned Scrams with Complications

a. Inspection Scope

The inspectors sampled licensee submittals for the Unplanned Scrams with Complications PI for the period from January 2013 to December 2013. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, were used. The inspectors reviewed the licensees operator narrative logs, CRs, event reports and NRC Integrated Inspection reports for the period to validate the accuracy of the submittals. The inspectors also reviewed the licensees CAP to determine if any problems had been identified with the PI data collected or transmitted for this indicator. Documents reviewed are listed in the to this report.

This inspection constituted one Unplanned Scrams with Complications sample as defined in IP 71151-05.

b. Findings

No findings were identified.

.3 Unplanned Transients per 7000 Critical Hours

a. Inspection Scope

The inspectors sampled licensee submittals for the Unplanned Transients per 7000 Critical Hours PI for the period from January 2013 through December 2013. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, were used. The inspectors reviewed the licensees operator narrative logs, CRs, maintenance rule records, event reports and NRC integrated IRs for the period to validate the accuracy of the submittals.

The inspectors also reviewed the licensees CAP to determine if any problems had been identified with the PI data collected or transmitted for this indicator. Documents reviewed are listed in the Attachment to this report.

This inspection constituted one unplanned transients per 7000 critical hours sample as defined in IP 71151-05.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

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

a. Inspection Scope

As part of the various base-line IPs discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that 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 CR 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 Follow-Up Sample for In-Depth Review: Licensee Corrective Actions for Adverse Trend

in Refueling Outage Industrial Safety Associated with Work from Elevated Heights

a. Inspection Scope

The inspectors performed a review of the licensees CAP and associated documents to assess the adequacy of corrective actions taken in response to an adverse trend in industrial safety performance that was identified during the early part of the licensee's 18th RFO, and in the preparation weeks immediately preceding the start of the outage on February 1, 2014.

The inspectors review was predominantly focused on the results of daily CAP item screenings discussed in Section 4OA2.2 above, but also included the review of specific materials (i.e., human performance stand down packages, special plant communications/event news flashes, etc.) developed by the licensee to address the adverse trend.

This review constituted a single follow-up inspection sample for in-depth review as defined in IP 71152-05.

b. Observations During the course of the inspection period, the inspectors noted an increasing trend in the incidence of industrial safety incidents and near-miss events. Many of these events were related to the common theme of industrial safety associated with personnel working at elevated heights. Specific examples associated with this trend included, but were not limited to:

• January 14, 2014. Inspectors conducting a routine plant status tour identified several scaffold builders working at an elevated height and not properly utilizing fall protection equipment, as required by the licensee's Generation Personal Safety Manual. The arrangement of the fall protection equipment being utilized was such that if a worker fell from the elevated height, they would have contacted the ground before their fall arrest equipment would have had an effect. After being informed of the inspectors' observation, the licensee modified the workers' fall protection equipment arrangement to include retractable lanyards, which alleviated the concern (CR 2014-00631);

• February 4, 2014. A welder received a cut to an elbow and required medical attention after an approximate 10 foot fall while transitioning from a ladder to a scaffold platform (CR 2014-02119);

• February 5, 2014. The grating to the containment normal sump was found removed, leaving an open hole in the containment floor open, unmarked, and unprotected (CR 2014-02125);

• February 5, 2014. An individual was observed in containment working at an elevated height without fall protection. The individual had accessed a scaffold that was tagged with a "Yellow" scaffold inspection tag, and the amplifying instructions on the tag required the use of fall protection equipment. When questioned, the individual stated that they had not read the scaffold tag prior to accessing the scaffold (CR 2014-02218); and

• February 5, 2014. An opening to the reactor refueling cavity in containment was found unattended and unmarked by a safety observer during a tour. The unguarded opening could have resulted in an individual falling some 30 feet into the reactor refueling cavity below (CR 2014-02225).

On February 6, 2014, the licensee generated a CR to document this adverse trend (CR 2014-02290). Immediate corrective actions taken by the licensee included specific discussion of the fall that occurred on February 4th in materials reviewed during a periodic stand down for all site personnel that took place during the period of February 5-6, 2014, as well as the inclusion of industrial safety requirements for personnel working at heights into the daily outage safety message covered with all personnel during shift turnovers during the period of February 6-7, 2014. Later actions taken by the licensee included the establishment of a "daily safety advocate" for each work crew prior to their shift, as well as increased manager/supervisor field observations focusing on compliance with safety standards. Despite these actions, however, similar industrial safety issues with personnel working at elevated heights continued to be observed:

• February 8, 2014. A scaffold crew working in containment lost control of a 6 foot metal scaffold pan, which fell approximately 25 feet to the floor below. An individual was struck on the arms by the falling object as it went down past his elevation (CR 2014-02446 and CR 2014-02447);

• February 8, 2014. A floor grating was found removed with only one side of the resultant opening being barricaded with red safety/danger rope, and the opening was not attended (CR 2014-02402);

• February 11, 2014. A piece of steel floor grating was found removed in containment creating an opening for an individual to fall approximately 20 feet to the elevation below (CR 2014-02593);

• February 13, 2014. An individual received a minor injury when, while attempting to view details associated with some overhead piping, their foot slipped off the edge of the scaffold. Corrective actions taken by the licensee after the event included installation of scaffold toe boards to provide protection from the edge of the scaffold decking (CR 2014-02808);

• February 14, 2014. A section of aluminum diamond deck plate fell from an upper elevation in containment. Although licensee and contractor personnel who witnessed the object fall and strike the ground conducted an immediate search as to the source from where the object fell, no source or responsible individual for the dropped item were ever identified (CR 2014-02901); and

• February 14, 2014. A 4 foot scaffold pole was dropped in the west condenser pit in the turbine building (CR 2014-02986).

On February 16, 2014, the licensee generated another CR to document the adverse trend being observed with dropped objects from heights (CR 2014-03057). Ultimately, the licensee closed this CR to actions already taken and an increased awareness on the part of site personnel for using lanyards and other devices to capture tools when working at elevated heights. Later that same day, an individual working in containment received multiple injuries in fall from ladder in containment. The individual was transported to a local hospital as a potentially contaminated injured person, and was accompanied by RP technicians to the hospital to facilitate decontamination efforts. The licensee formally reported this event to the NRC Headquarters Operations Officer in accordance with 10 CFR 50.72 (b)(3)(xii); further details can be found in Section 4OA3.3 of this report.

The licensee entered the event into their CAP as CR 2014-03079.

Following the personnel injury event discussed above, the site experienced no additional serious industrial safety events associated with work at elevated heights. However, non-consequential near-miss issues, particularly with objects dropped while working from elevated heights, have continued. Albeit their frequency and severity has been diminished, these events remain a challenge for licensee management and craft personnel.

c. Findings

No findings were identified.

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

.1 Multiple Fire Alarms in the Boric Acid Tank Room

a. Inspection Scope

The inspectors reviewed the licensees response to the control room receiving multiple fire alarms in the boric acid addition tank room located in the auxiliary building on February 13, 2014. The licensee entered their abnormal operating procedure for fire and determined that a fault in one of the primary boric acid addition tank room heaters triggered three fire detectors to alarm. No actual fire or indications of an explosion were observed. The inspectors reviewed the licensees EP to verify none of the emergency action level entry criteria were met, verified operability of the boric acid addition tanks (room temperature) was maintained, and reviewed the licensees corrective actions to address the faulted room heater. Documents reviewed in this inspection are listed in the

.

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

b. Findings

No findings were identified.

.2 Concrete Void Identified in the Reactor Shield Building: Event Notification 49828

a. Inspection Scope

The inspectors reviewed the licensees response to an unanalyzed condition identified on February 14, 2014, with the plant offline and the reactor defueled. During hydrodemolition operations to cut an access opening in the shield building to facilitate steam generator replacement, the licensee identified a void in the concrete at the top of the area which was restored following reactor vessel head replacement in 2011. The licensee reported the condition to the NRC per 10 CFR 50.72(b)(3)(ii)(B). On February 24, 2014, the licensee completed a technical evaluation of the condition and concluded that the void in the concrete had not rendered the shield building incapable of performing any of its required functions specified in the USAR. The licensee subsequently retracted their 8-hour non-emergency notification (Event Number 49828),which had previously been made to the NRC under 10 CFR 50.72(b)(3)(ii)(B).

The inspectors reviewed the licensees response to the condition and their decision to report the issue to the NRC under 10 CFR 50.72(b)(3)(ii)(B), as well as their decision to retract that report.

In addition to standard RFO activities, activities during the site's 18th RFO also included replacement of the unit's two steam generators and significant portions of the connected RCS piping. Inspection of the steam generator replacement and associated activities, including the NRC's technical review of the Shield Building concrete void condition discussed in this section, is covered under NRC IP 50001, "Steam Generator Replacement Inspection," and will be documented in a separate NRC Inspection Report 05000346/2013010. Applicable portions of the inspectors' reviews documented in this section also were credited toward completion of IP 50001. Documents reviewed are listed in the Attachment to this report.

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

b. Findings

No findings were identified.

.3 Potentially Contaminated Person Requiring Offsite Medical Attention: Event Notification

49833

a. Inspection Scope

The inspectors reviewed the licensees response to a medical emergency requiring offsite assistance on February 16, 2014. While working in the reactor containment building supporting the installation of temporary radiation shielding, an individual fell approximately eleven feet to the 565 elevation floor while descending on a permanently installed plant ladder. The individual received head and multiple bodily injuries and was required to be transferred to a hospital via ambulance. Because the individual had been working in a potentially contaminated area and was wearing anti-contamination clothing, site RP technicians accompanied the individual to the hospital. The licensee reported the condition to the NRC per 10 CFR 50.72(b)(3)(xii), as a medical emergency involving a potentially contaminated individual and requiring offsite medical assistance.

The inspectors reviewed the licensees response to the event, including the onsite response, emergency transport via ambulance, and the licensee's response activities taken at the hospital. Additionally, the inspectors reviewed the licensee's decision to report the issue to the NRC under 10 CFR 50.72(b)(3)(xii).

The inspectors reviewed the licensees radiological surveys (14-6212, 14-6213) and the associated Personnel Decontamination Form (DBEP 059-02), Ambulance Radiological Release Survey Form (DBEP 060-01), and the Medical Facility Radiological Release Form (DBEP 061-01) generated as a result of the personnel injury on February 16, 2014. The surveys were conducted by qualified personnel using calibrated and operable radiation detection instrumentation, and the surveys were appropriately reviewed by RP Supervisors. The individual was not identified as contaminated.

However, the protective clothing worn by the individual in containment was identified as contaminated at low levels. The protective clothing and all potentially contaminated disposable medical supplies used by the medical staff were retrieved and returned to the plant for appropriate disposition. Surveys indicated that the medical facility and ambulance were free of contamination following transport and care of the injured person.

Ambulance personnel were assigned thermoluminescent and electronic dosimeters.

Dosimeters assigned to the transport personnel did not show any measurable radiation.

Surveys noted the dose rates in the ambulance during transport were not above ambient background. Documents reviewed are listed in the Attachment to this report.

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

b. Findings

No findings were identified.

4OA5 Other Activities

.1 Historical Findings Cross-Reference for Cross-Cutting Aspects

The table below provides a cross-reference from the 3rd and 4th Quarter 2013 findings and associated cross-cutting aspects to the new cross-cutting aspects resulting from the common language initiative. These aspects and any others identified since January 2014 will be evaluated for cross-cutting themes and potential substantive cross-cutting issues in accordance with IMC 0305 starting with the 2014 mid-cycle assessment review.

Finding Old Cross-Cutting Aspect New Cross-Cutting Aspect 05000346/2013004-01 H.1(c) H.10 05000346/2013004-02 P.1(c) P.2 05000346/2013004-03 H.2(d) H.1 05000346/2013004-04 P.1(c) P.2 05000346/2013005-01 H.4(a) H.12 05000346/2013005-02 H.3(a) H.5

4OA6 Management Meetings

.1 Exit Meeting Summary

On April 15, 2014, the inspectors presented the inspection results to Mr. R. Lieb, the Site Vice President, 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. Proprietary material received during the inspection was returned to the licensee.

.2 Interim Exit Meetings

Interim exits were conducted for:

• the inspection results for the areas of radiological hazard assessment and exposure controls; occupational ALARA planning and controls; in-plant airborne radioactivity control and mitigation; and radioactive solid waste processing and radioactive material handling, storage, and transportation with Mr. B. Boles, the Director of Site Operations, on March 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.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

R. Lieb, Site Vice President

B. Boles, Director, Site Operations

K. Byrd, Director, Site Engineering

G. Cramer, Manager, Site Protection

J. Cuff, Manager, Training

J. Cunnings, Manager, Site Maintenance

A. Dawson, Manager, Chemistry

D. Hartnett, Superintendent, Operations Training

J. Hook, Manager, Design Engineering

D. Imlay, Director, Site Performance Improvement

G. Kendrick, Manager, Site Outage Management

B. Kremer, Manager, Plant Engineering

P. McCloskey, Manager, Site Regulatory Compliance

D. Noble, Manager, Radiation Protection

W. OMalley, Manager, Nuclear Oversight

R. Oesterle, Superintendent, Nuclear Operations

R. Patrick, Manager, Site Work Management

D. Petro, Manager, Steam Generator Replacement Project

T. Summers, Manager, Site Operations

M. Roelant, Manager, Site Projects

L. Rushing, Director, Special Projects

D. Saltz, Director, Site Maintenance

J. Sturdavant, Regulatory Compliance

L. Thomas, Manager, Nuclear Supply Chain

M. Travis, Superintendent, Radiation Protection

J. Vetter, Manager, Emergency Response

G. Wolf, Supervisor, Regulatory Compliance

K. Zellers, Supervisor, Reactor Engineering

LIST OF ITEMS

Opened

05000346/2014002-01 NCV Failure to Implement Fire Protection Plan Requirements Related to Control of Ignition Sources (Section 1R05.1)

05000346/2014002-02 NCV Containment ECCS Recirculation Sump Reliability Degraded Due to Unfastened Debris Gate (Section 1R20.1)

Closed

05000346/2014002-01 NCV Failure to Implement Fire Protection Plan Requirements Related to Control of Ignition Sources (Section 1R05.1)

05000346/2014002-02 NCV Containment ECCS Recirculation Sump Reliability Degraded Due to Unfastened Debris Gate (Section 1R20.1)

Discussed

None