IR 05000247/2011002

IR 05000247-11-002; 01/01/2011-03/31/2011; Indian Point Nuclear Generating (Indian Point) Unit 2; Equipment Alignment and Post-Maintenance Testing

ML111320196
Person / Time
Site:	Indian Point
Issue date:	05/12/2011
From:	Mel Gray Reactor Projects Branch 2
To:	Joseph E Pollock Entergy Nuclear Operations
Gray, Mel NRC/RGNI/DRP/PB2/610-337-5209
References
Download: ML111320196 (50)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION I

475 ALLENDALE ROAD KlNG OF PRUSSIA. PA r9406-1415

+***+ May 12, ZOIL Mr. Joseph Site Vice President Entergy Nuclear Operations, lnc.

Indian Point Energy Center 450 Broadway, GSB Buchanan. NY 1051 1-0249 SUBJECT: INDIAN POINT NUCLEAR GENERATING UNIT 2 - NRC INTEGRATED I NSPECTTON REPORT 05000247 l2A1 1 002

Dear Mr. Pollock:

On March 31,2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at f ndian Point Nuclear Generating Unit 2. The enclosed integrated inspection report documents the inspection results, which were discussed on April 18, 2011 with you and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents two NRC-identified findings of very low safety significance (Green).

These findings were determined to involve violations of NRC requirements. However, because of their very low safety significance and because they are entered into your corrective action program (CAP), the NRC is treating these as non-cited violations (NCVs) consistent with Section 2.3.2 of the NRC Enforcement Policy. lf you contest these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region 1; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Senior Resident Inspector at Indian Point Nuclear Generating Unit 2. ln addition, if you disagree with the cross-cutting aspect assigned to the findings 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 1, and the NRC Senior Resident lnspector at Indian Point Nuclear Generating Unit 2. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room of from the Publicly Available Records component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web Site at http://www.nrc.sov/readinq-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

,7 Mel Gray, Chief Projects Branch Division of Reactor Projects Docket No. 50-247 License No. DPR-26 Enclosure: lnspection Report No. 05000247 12011002 M Attachment: Supplemental Information cc w/encl: Distribution via ListServ

SUMMARY OF FINDINGS

IR 05000247/2011002; 01/01/2011 - 03/31/2011; Indian Point Nuclear Generating (Indian

Point) Unit 2; Equipment Alignment and Post-Maintenance Testing.

This report covered a three-month period of inspection by resident and region based inspectors.

Two NCVs of very low safety significance (Green) were identified. These findings were also determined to be violations of NRC requirements. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,

Significance Determination Process (SDP). The cross-cutting aspects for the findings were determined using IMC 0310, Components within the Cross-Cutting Areas. Findings for which the SDP does not apply may be Green, or be assigned a severity level after NRC management review. The NRCs program for overseeing safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings, because Entergy procedure 2-COL-18.1,

Main Steam and Reheat System, was not adequate to ensure closure of main steam isolation valve (MSIV) bypass stop valve MS-55D. Specifically, between April 10, 2010 and September 12, 2010, procedure 2-COL-18.1 did not provide adequate instructions to operators to ensure MS-55D was closed, which resulted in MS-55D being left partially open, and unable to isolate the 24 steam generator (SG) during accident conditions.

Entergy personnel took immediate corrective actions to close MS-55D. This issue was entered into Entergys CAP as condition reports (CRs) IP2-2010-05694 and IP2-2010-06745.

This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the inadequate procedure resulted in the manual 3-inch MSIV bypass stop valve MS-55D for the 24 SG being left partially open for approximately five months. Based on NRC senior reactor analyst review, it was determined that operators could have isolated the other three SGs with their MSIVs and steamed them to remove decay heat and depressurize the plant using their atmospheric dump valves, while isolating the 24 SG further down the main steam system at the turbine bypass and stop valves. Therefore, using IMC 0609.04,

"Phase 1 - Initial Screening and Characterization of Findings," the inspectors determined this finding was of very low safety significance (Green) because the finding did not result in a loss of the safety function given the operators ability to isolate the other SGs and the 24 SG with the turbine bypass and stop valves. Additionally, the finding was not potentially risk significant due to a seismic, flooding, or severe weather initiating event.

The inspectors determined there was no cross-cutting issue associated with the finding because the performance deficiency did not reflect Entergy's current performance.

Specifically, the procedure change occurred more than three years ago and was outside the current assessment period. (Section 1R04)

Green.

The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI,

Corrective Action, because Entergy personnel did not promptly identify and correct an adverse condition related to a service water (SW) pipe leak. Specifically, on October 29, 2010, NRC inspectors identified a leak on the base weld of the 25 SW pipe vacuum breaker which required subsequent evaluation and repair by Entergy personnel to restore operability of the 25 service water pump (SWP). This issue was entered into Entergys CAP as CR IP2-2010-6620.

This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the 25 SW pipe weld leak challenged the capability and the reliability of the SWP, and the pump was declared inoperable by Entergy personnel to conduct repairs. Using IMC 0609.04,

Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because the finding was not related to a design or qualification deficiency, did not represent a loss of system safety function, and the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.

The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the CAP attribute because Entergy personnel did not implement a CAP with a low threshold for identifying issues, specifically, identifying a leak on the 25 SWP piping. P.1(a) per IMC 0310] (Section 1R19)

REPORT DETAILS

Summary of Plant Status

Indian Point Unit 2 began the inspection period operating at full reactor power (100%) and remained at or near full power during the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

The inspectors performed a review of Entergy procedures to address seasonal cold weather conditions. This review included an evaluation of deficiencies identified by Entergy personnel during the current seasonal preparations, and that adverse conditions were being adequately addressed to ensure the cold weather conditions would not have significant impact on plant operation and safety. The inspectors conducted plant and system walkdowns of the refueling water storage tank, the auxiliary feedwater building, SW intake structure, and the control building. Additionally, the inspectors conducted the review to verify that the stations implementation of OAP-008, "Severe Weather Preparations," and OAP-048, "Seasonal Weather Preparation," appropriately maintained systems required for normal operation and safe shutdown conditions.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one sample as defined in NRC Inspection Procedure (IP) 71111.01.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial System Walkdowns

a. Inspection Scope

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

January 19, 2011, 21 containment spray (CS) pump after breaker replacement; March 15, 2011, 24 main steam line after November outage; and March 15, 2011, 13.8 kV circuit after a loss of the 138 kV circuit.

The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors focused on those conditions that could affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, updated final safety evaluation report (UFSAR), TSs, work orders (WOs),

CRs, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have impacted system performance of their intended safety functions. The inspectors also inspected 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 deficiencies. The inspectors also reviewed whether Entergy staff had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the CAP with the appropriate significance characterization.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of three partial system walkdown samples as defined in NRC IP 71111.04.

b. Findings

Main Steam System Configuration Control Procedure not Adequate to Ensure Closure of MS-55D

Introduction:

The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy procedure 2-COL-18.1, Main Steam and Reheat System, was not adequate to ensure closure of MSIV bypass stop valve MS-55D. Specifically, between April 10, 2010 and September 12, 2010, procedure 2-COL-18.1 did not provide adequate instructions to operators to ensure MS-55D was closed, which resulted in MS-55D being left partially open.

Description:

On September 12, 2010, during plant heat up coming out of a maintenance outage, operations personnel were not able to obtain the desired pressurization of the main steam line from the 24 SG. During investigation Entergy personnel determined MSIV bypass stop valve MS-55D was not fully closed as required by 2-COL-18.1, Main Steam and Reheat System. Operations personnel tried to close the manual valve using the reach rod, but no movement was observed. Entergy staff then applied a torque amplifying device, and the valve closed one additional turn. Maintenance personnel lubricated the valve stem and the valve closed approximately four more turns.

The main steam line from the 24 SG has a MSIV MS 1-24, and MSIV bypass stop valve MS-55D, that are required to close or be closed during a design basis accident. Valve MS-55D is manually opened during plant startup to warm the main steam header and to equalize the pressure on either side of MS 1-24. When no longer needed for steam line warming, valve MS-55D is manually closed and is required to remain closed while the plant is operating. Valve MS-55D is required to be closed during plant operation to mitigate the following postulated accident scenarios: main steam line break, locked rotor of reactor coolant pump (RCP), small break loss-of-coolant accident, rod ejection accident, and steam generator tube rupture (SGTR).

Entergy personnel initiated CR-IP2-2010-05694 to address valve MS-55D being identified partially open out of position and took corrective actions to increase the lubrication preventative maintenance (PM) from four years to two years. Entergy staff determined valve MS-55D was last operated on April 10, 2010, when Unit 2 was starting up from refueling outage 2R19, and had been left partially open since then. However, Entergy staff determined this valve being left partially open was not a maintenance rule functional failure and that the valve would have performed its safety function during a design basis accident.

The inspectors reviewed procedure 2-COL-18.1 and walked down valve MS-55D. The inspectors determined that valve MS-55D had no control room or local position indication for operations personnel to determine if the valve is closed. The inspectors determined 2-COL-18.1 did not provide guidance to aid in operations personnel determining if valve MS-55D is closed. Further, the inspectors determined procedure 2-COL-18.1 did not provide adequate guidance to ensure MS-55D was closed considering that there was no control room or local valve position indication.

The inspectors reviewed CR-IP2-2010-05694 to assess whether valve MS-55D would have performed its safety function to mitigate an accident, because the valve was left open approximately five turns and required lubrication to close the valve fully. The inspectors determined that during an accident with this valve partially open, that operations personnel would not have been able to isolate the 24 SG with valve MS-55D if the 24 SG was subject to a steam line break or tube rupture condition. Based on inspector questions, Entergy personnel re-evaluated the maintenance rule functional failure determination and subsequently determined that the valve being left partially open was a maintenance rule functional failure and would not have been able to perform its function to be closed during a design basis accident.

The inspectors also determined that for the period this valve was partially open, that the dose consequences for design basis accidents were not addressed in Entergys CR.

Entergy personnel wrote CR-IP2-2010-06745 to address the dose consequence concerns raised by the inspectors. Entergy personnel evaluated the dose consequences and determined that the current accident analysis bounded this valve being partially open.

Analysis:

The performance deficiency associated with this finding was that Entergy procedure 2-COL-18.1 was not adequate to ensure MSIV bypass stop valve MS-55D was closed on April 10, 2010. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage).

Specifically, the inadequate procedure resulted in the manual 3-inch MSIV bypass stop valve MS-55D for the 24 SG being left partially open for approximately five months.

Neither the MSIVs nor their bypass valves are directly modeled in either the Phase 2 or the Phase 3 SDP risk evaluations tools; therefore, a conservative analysis was conducted to encompass the impact of this partially open 3-inch manual valve. Based on NRC senior reactor analyst review, if it was assumed that the partially open MS-55D would have been equivalent to the 24 SG MSIV not closing following a 24 SGTR, the operators, in accordance with emergency operating procedures, could have isolated the other three SGs with their MSIVs and steamed them to remove decay heat and depressurize the plant using their atmospheric dump valves while isolating the 24 SG further down the main steam system at the turbine bypass and stop valves. Therefore, using IMC 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," the inspectors determined this finding was of very low safety significance (Green) because the finding did not result in a loss of the safety function given the operators ability to isolate the other SGs and the 24 SG with the turbine bypass and stop valves.

Additionally, the finding was not potentially risk significant due to a seismic, flooding, or severe weather initiating event.

The inspectors determined there was no cross-cutting issue associated with the finding because the performance deficiency did not reflect Entergy's current performance.

Specifically, the procedure change occurred more than three years ago and was outside the current assessment period.

Enforcement:

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

Procedure 2-COL-18.1, Main Steam and Reheat System, required that MS-55D be closed and locked. Contrary to the above, Entergy procedure 2-COL-18.1 was not adequate to ensure MS-55D was closed by operators, which resulted in MS-55D being left partially open between April 10, 2010 and September 12, 2010, and unable to provide its safety function to isolate the 24 SG during accident conditions. Entergy personnel took immediate corrective actions to close MS-55D. Because the violation was of very low safety significance and it was entered into Entergys CAP as IP2-2010-05694 and IP2-2010-06745, this violation is being treated as a NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000247/2011002-01, Main Steam System Configuration Control Procedure not Adequate to Ensure Closure of MS-55D.

.2 Full System Walkdown

a. Inspection Scope

On March 7 and 8, 2011, the inspectors performed a complete system alignment inspection of the emergency diesel generators (EDGs) to verify the functional capability of the system. The inspectors selected this system because it was considered both safety significant and risk significant in Entergys probabilistic risk assessment. The inspectors inspected the system to review mechanical and electrical equipment line ups, electrical power availability, component lubrication, and equipment cooling, fuel oil supply, hanger and support functionality, operability of support systems, and to ensure that ancillary equipment or debris did not interfere with equipment operation. In addition, the inspectors reviewed the CAP database to ensure that system adverse conditions were being identified and appropriately resolved.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one complete system walkdown sample as defined in NRC IP 71111.04.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Resident Inspector Quarterly Walkdowns

a. Inspection Scope

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

Pre-Fire Plan (PFP) 252; PFP-255A; PFP-255B; PFP-255E; PFP-255C; and PFP-255D.

The inspectors reviewed areas to assess if Entergy personnel 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 stations fire plan. The inspectors selected fire areas based on their overall contribution to internal fire risk and their potential to affect equipment that could initiate or mitigate a plant transient. Using the documents listed in the attachment, the inspectors reviewed whether 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 that fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also reviewed whether issues identified during the inspection were entered into the licensees CAP.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of six quarterly fire protection inspection samples as defined in NRC IP 71111.05.

b. Findings

No findings were identified.

.2 Annual Fire Drill

a. Inspection Scope

On February 10, 2011, the inspectors observed a fire brigade activation involving a simulated fire in the vicinity of the Appendix R / Station Black Out EDG, which is located in the turbine building. The observation involved an evaluation of the readiness of the plant fire brigade to fight fires. The inspectors reviewed whether Entergy staff identified performance deficiencies; openly discussed them in a critical manner at the drill debrief; and identified appropriate corrective actions. Specific attributes evaluated by the inspectors were

(1) proper wearing of turnout gear and self contained breathing apparatus;

(2) proper use and layout of fire hoses;

(3) employment of appropriate fire fighting techniques;

(4) sufficient firefighting equipment brought to the scene;

(5) effectiveness of fire brigade leader communications, command, and control;

(6) search for victims and propagation of the fire into other plant areas;

(7) smoke removal operations;

(8) utilization of preplanned strategies;

(9) adherence to the preplanned drill scenario; and

(10) drill objectives.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one annual fire protection inspection sample as defined in NRC IP 71111.05.

b. Findings

No findings were identified.

1R06 Flood Protection Measures

Cables Located in Underground Manholes Inspection

a. Inspection Scope

The inspectors performed an inspection of underground Manhole 21, which contains safety related electrical cabling to the SWPs. The inspectors reviewed the UFSAR and related design basis documents to identify the requirements for the manhole design. The inspectors assessed the material condition of the support trays and cable insulation to verify there was no evidence of conditions that could challenge operability of the safety related pumps. The inspectors reviewed whether adverse conditions discovered during the manhole inspection, if applicable, were entered into Entergy's CAP.

Specific documents reviewed during this inspection are listed in the attachment. This inspection completes one of the two required manhole inspections in accordance with NRC IP 71111.06.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

Quarterly Review

a. Inspection Scope

On January 25, 2011, the inspectors observed a crew of licensed operators, responding to a simulated event involving a small break loss of coolant accident resulting in degraded core cooling and the failure of select components to automatically start as required. The inspectors observed the scenario in the plant simulator to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and that training was being conducted in accordance with station procedures. The inspectors evaluated the following areas regarding crew and operator performance:

Clarity and formality of communications; Implementation of timely actions; Prioritization, evaluation, and verification of annunciator alarms; Usage and implementation of abnormal and emergency procedures; Control board operations; Identification and implementation of TS actions and emergency plan actions and notifications; and Oversight and direction from control room supervisors.

The inspectors compared the crews performance in these areas to critical task completion requirements.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one quarterly licensed operator requalification program sample as defined in NRC IP 71111.11.

b. Findings

URI 05000247/2011002 - Notification Process for State/Local Authorities During a Simulator Scenario

Introduction:

Based on a simulator drill scenario on January 25, 2011, the inspectors identified an issue of concern regarding whether Entergy procedure IP-EP-210, "Central Control Room." Attachment 9.1, Shift Manager/Plant Operations Manager (Emergency Director) Checklist, is adequate to ensure proper notification of state and local authorities as required by IPEC Emergency Plan Section E. Additionally, the inspectors questioned whether operator training with regard to implementation of this procedure checklist is adequate and consistent amongst operator crews. As a result, the NRC has opened an unresolved item (URI) requiring further information from Entergy regarding their review of the adequacy of the procedure including an assessment of operator training specific to implementation of that procedure checklist.

Description:

Following the emergency declaration of an Alert by operators during a simulator drill scenario on January 25, 2011, the operators entered emergency plan implementing procedure IP-EP-210, "Central Control Room," Attachment 9.1, Shift Manager/Plant Operations Manager (Emergency Director) Checklist. The IPEC Emergency Plan, Section E, Notification Methods and Procedures, paragraph 1.b.5, requires in part that an immediate notification (within 15 minutes) of an Alert is made by the Shift Manager or his designee to the New York State and Westchester, Rockland, Putnam, and Orange Counties. The emergency plan implementing procedure checklist directs the Shift Manager to complete a New York State (NYS) Radiological Emergency Data Form and have a control room Offsite Communicator email and fax the data form to the offsite authorities. The Offsite Communicator must then confirm receipt of the information by offsite authorities. NRC regulations, specifically 10 CFR 50.47(b)(5),require in part that "procedures have been established for notification, by the licensee, of State and local response organizations."

The drill scenario simulated one county not being present during the initial notification call via the radiological emergency communication system (RECS). The Offsite Communicator provided the event notification to NYS and the counties that were present on the line. The NRC inspectors observed that during the drill the Offsite Communicator did not implement additional communication measures to ensure the county, not present during the initial notification, received the event notification via fax. The inspectors observed that not affirming receipt of the notification by the county would not be consistent with IPEC Emergency Plan Section E in ensuring the licensee notifies all state and local authorities. The inspectors also observed that Entergy evaluators did not address this issue during the simulator scenario critique. The inspectors questioned Entergy personnel regarding their views during the simulator scenario and the expected operator response. The inspectors concluded additional information is required from Entergy staff related to their assessment regarding the adequacy of the procedure IP-EP-210, Attachment 9.1 and operator training with regard to the implementation of that procedure. Prior to completion of this inspection, Entergy personnel revised the Control Room Initial Notification Checklist (Form EP-4) to provide direction to operators in the event initial notifications are not able to be completed for required state and local authorities. (URI 05000247/2011002-02, Notification Process for State/Local Authorities During a Simulator Scenario)

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the following areas to assess the effectiveness of maintenance activities on system performance and reliability. The inspectors reviewed, when applicable, system health reports, CAP documents, maintenance WOs, and maintenance rule basis documents to ensure performance problems were being identified and properly evaluated within the scope of the maintenance rule. For each sample selected, the inspectors reviewed whether the structure, system, and component (SSC) was properly scoped into the maintenance rule in accordance with 10 CFR 50.65 and reviewed whether the (a)(2) performance criteria established by Entergy staff were appropriate. For SSCs classified as (a)(1), the inspectors assessed the adequacy of goals and corrective actions to return these SSCs to (a)(2). Additionally, the inspectors determined if Entergy staff was identifying and addressing common cause failures that occurred within and across maintenance rule system boundaries.

Dual indication on 22 fan cooler unit coil outboard inlet stop valve SWN- 41-2B; and Chemical and volume control system charging flow from regenerative heat exchanger 21 to Loop 22 hot leg check valve 210A failed to stroke closed.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of two quarterly maintenance effectiveness sample as defined in NRC IP 71111.12.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed station 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:

January 5, 2011, with control rods in manual due to 480V undervoltage planned testing; January 31, 2011, with 24 rectifier, 21 instrument air dryer, and 22 charging pump out of service for unplanned maintenance and planned safety injection logic testing; February 1, 2011, with 24 rectifier, 21 instrument air dryer, and 22 charging pump out of service for unplanned maintenance, severe weather, and planned steam flow / feed flow mismatch testing; February 22, 2011; with 24 rectifier, 23 SWP, and PCV-1139 out of service for unplanned maintenance; and March 1, 2011, with loss of the normal offsite power 138 kV circuit and the back-up 13.8 kV circuit manually placed in service.

The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that Entergy personnel performed risk assessments as required by 10 CFR 50.65(a)(4)and that the assessments were accurate and complete. When Entergy personnel performed emergent work, the inspectors reviewed whether operations personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work and discussed the results of the assessment with the stations probabilistic risk analyst or shift technical advisor, to verify plant conditions were consistent with the risk assessment. The inspectors also reviewed the TS requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of five maintenance risk assessments and emergent work control inspection sample as defined in NRC IP 71111.13.

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

January 3, 2011, 23 SG feed flow indicator FT-438B reading erratically; January 6, 2011, impact of unplanned control rod movement on Tavg; January 13, 2011, 22 RCP seal water delta temperature less than other RCPs; February 6, 2011, 22 SG alternate safe shutdown level indicator LI-5002-1 reading high; and March 2, 2011, 21 EDG jacket water pressure switch high resistance.

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 assess whether 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 TSs and UFSAR to Entergys 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 also reviewed a sampling of corrective action documents to verify that Entergy personnel were identifying and correcting any deficiencies associated with operability evaluations.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of five operability evaluations inspection samples as defined in NRC IP 71111.15.

b. Findings

No findings were identified.

1R18 Plant Modifications

.1 Temporary Modifications

a. Inspection Scope

The inspectors reviewed the following temporary modification to verify that the safety functions of affected safety systems were not degraded:

On April 6, 2010, Entergy staff implemented WO 231016 to block isolation valve 204A closed. Chemical and volume control system charging flow from regenerative heat exchanger 21 to Loop 22 hot leg check valve 210A was not functioning properly, and system design requires two valves in series to maintain reactor coolant system boundary integrity in the case of an accident. Valve 204A was blocked closed to provide for the two valves in series boundary. The modification is planned to be removed during the next refueling outage.

The inspectors reviewed the temporary modification and the associated safety evaluation screening against the system design bases documentation, including the UFSAR and the TSs, to verify that the modification did not adversely affect the system operability/availability. The inspectors also reviewed whether the installation and restoration were consistent with the modification documents and that configuration control was adequate. Additionally, the inspectors reviewed whether the temporary modification was identified on control room drawings, appropriate tags were placed on the affected equipment, and Entergy personnel evaluated the combined effects on mitigating systems and the integrity of radiological barriers.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one sample for temporary plant modifications as defined in NRC IP 71111.18.

b. Findings

No findings were identified.

.2 Permanent Modifications

a. Inspection Scope

The inspectors reviewed the following permanent modification to verify that the safety functions of affected safety systems were not degraded:

On January 13, 2011, Entergy staff implemented WO 250423 on the 23 RCP to increase the vibration setpoints from providing an alarm indication at 12 and 15 mils to providing an alarm indication at 15 and 20 mils. Work orders 224117, 224118, and 224119 were written to change the setpoints for 21, 22, and 24 RCPs.

The inspectors reviewed the permanent modification and the associated safety evaluation screening against the system design bases documentation, including the UFSAR and the TSs to verify that the modification did not adversely affect the system operability and/or availability. The inspectors also reviewed whether the installation and restoration were consistent with the modification documents and that configuration control was adequate. Additionally, the inspectors reviewed whether the permanent modification was identified on control room drawings, appropriate tags were placed on the affected equipment, and Entergy personnel evaluated the combined effects on mitigating systems and the integrity of radiological barriers.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one sample for permanent plant modifications as defined in NRC IP 71111.18.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

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

November 25, 2010, 25 SWP after weld leak repair; January 13, 2011, 24 fan cooler unit vibration testing after fan bearing housing replacement; January 19, 2011, 21 CS pump after breaker replacement; January 29, 2011, 21 SG level indicator after associated bistable replacement; February 22, 2011, 22 auxiliary feedwater pump steam supply valve after valve positioner work; February 25, 2011, 23 SWP after maintenance including vacuum breaker replacement ; and March 21, 2011, 21 EDG after maintenance.

The inspectors selected these activities based upon the SSC's ability to affect risk. The inspectors evaluated these activities to determine (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; and that test instrumentation was appropriate. The inspectors evaluated the activities against the TSs, the UFSAR, 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 PMTs to determine whether Entergy personnel were identifying problems and entering them in the CAP and that the problems were being corrected commensurate with their importance to safety.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of seven PMT inspection samples as defined in NRC IP 71111.19.

b. Findings

Entergy Personnel Did Not Identify a Leak on the 25 SWP Piping

Introduction:

The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Entergy personnel did not promptly identify and correct an adverse condition related to a SW pipe leak. Specifically, on October 29, 2010, inspectors identified a leak on the base weld of the 25 SW pipe vacuum breaker which required subsequent evaluation and repair by Entergy personnel to restore operability of the 25 SWP.

Description:

On October 29, 2010, during a walkdown of the SWP pit, the inspectors identified a leak on the 25 SWP 14 discharge pipe vacuum breaker base weld and staining on the piping where the leak occurred. Entergy personnel initiated CR-IP2-2010-6620, performed an operability determination, but could not perform an ultrasonic test (UT) of the leakage area due to the irregular surface configuration of the weld. Entergy personnel determined the pipe was structurally sound; however, with the estimated hole size, the flow out of the pipe exceeded the capability of the strainer pit sump pump and could flood the pit, resulting in the inoperability of the SW system, so the pump was declared inoperable. The pipe was repaired with an ASME Section XI repair. A UT was performed on the removed section of pipe and found two rejectable linear indications, one was 3/8 in length, and the other, which was the source of the leak, was 1.25 in length.

The inspectors visually identified that the leak was occurring from the weld between the discharge piping and the vacuum breaker. The inspectors also concluded that the leak existed for at least a couple days based on engineering judgment with regard to observable pipe staining indications. The inspectors also noted this leak was at eye level and should have been readily visible to Entergy personnel entering the pit area for operator rounds.

The inspectors reviewed operations procedure OAP-017, Plant Surveillance and Operator Rounds, specifically step 1.2, which states Management expectations concerning rounds are operators should always be alert for the following: obvious leakage. Procedure EN-LI-102, Corrective Action Process, Step 5.2(e) states, in part, that Employees are required to initiate CRs for adverse conditions. Per EN-LI-102, an adverse condition is defined in part as A defect, characteristic, state or activity that prohibits or detracts from safe, efficient nuclear plant operation or a condition that could credibly impact nuclear safety, personnel safety, plant reliabilityAdverse conditions include non-conformances, conditions adverse to qualityExamples of adverse conditions are contained in Attachment 9.2. Attached 9.2 gives examples of an adverse condition including: Chemical or other leaks that could potentially impact plant operations or the environment. The inspectors determined operations personnel should have identified the SW leak consistent with expectations in procedures OAP-017 and EN-LI-102 to identify the stained piping and the leakage from the 25 SWP discharge piping and write a CR, so that the operability of the SW system could be evaluated.

Analysis:

The performance deficiency associated with this finding was that Entergy personnel did not promptly identify and correct an adverse condition related to a SW pipe leak. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the 25 SW pipe weld leak challenged the reliability and the capability of the SWP, and the pump was declared inoperable by Entergy personnel to conduct repairs. Using IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance (Green) because the finding was not related to a design or qualification deficiency, did not represent a loss of system safety function, and the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.

The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the CAP attribute because Entergy personnel did not implement a CAP with a low threshold for identifying issues specific to identification of a leak on the 25 SWP piping. P.1(a) per IMC 0310]

Enforcement:

10 CFR 50, Appendix B, Criterion XVI, Corrective Action Program, requires, in part, that the licensee assure that conditions adverse to quality, such as deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Entergy operations procedure OAP-017, Plant Surveillance and Operator Rounds, specifically step 1.2 states Management expectations concerning rounds are operators should always be alert for the following:

obvious leakage. Procedure EN-LI-102, Corrective Action Process, Step 5.2(e)states, in part, that Employees are required to initiate CRs for adverse conditions. Per EN-LI-102, an adverse condition is defined in part as A defect, characteristic, state or activity that prohibits or detracts from safe, efficient nuclear plant operation or a condition that could credibly impact nuclear safety, personnel safety, plant reliabilityAdverse conditions include non-conformances, conditions adverse to qualityExamples of adverse conditions are contained in Attachment 9.2. Attached 9.2 gives examples of an adverse condition including: Chemical or other leaks that could potentially impact plant operations or the environment. Contrary to the above, on October 29, 2010, inspectors identified a leak on the base weld of the 25 SW pipe vacuum breaker and staining on the piping where the leak occurred, which resulted in the inoperability of the 25 SWP.

Because this finding was of very low safety significance and was entered into Entergys CAP as CR-IP2-2010-6620, consistent with Section 2.3.2 of the Enforcement Policy, this violation is being treated as a NCV: NCV 05000247/2011002-03, Entergy Personnel Did Not Identify a Leak on the 25 Service Water Pump Piping.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed performance of surveillance tests and/or reviewed test data of selected risk significant SSCs, to assess whether test results met TSs, UFSAR, technical requirements manual, and Entergy procedure requirements. The inspectors reviewed whether the test acceptance criteria were sufficiently clear; tests demonstrated operational readiness and were consistent with design basis documentation; test instrumentation had accurate calibrations and appropriate range and accuracy for the application; tests were performed as written; and applicable test prerequisites were satisfied. Following the tests, the inspectors considered whether the test results supported conclusions that equipment was capable of performing the required safety functions. The following surveillance tests were reviewed:

January 4, 2011, 2-PT-A035C, 23 Station Battery Intercell Resistance Check; January 5, 2011, 2-PT-2Y008A, 21 EDG Mechanical Overspeed Trip; January 10, 2011, 2-PT-Q28A, 21 Residual Heat Removal Pump Inservice Test; January 31, 2011, 2-PT-2M4, Safety Injection System Train A Actuation Logic and Master Relay Test; March 8, 2011, 2-PT-Q013-DS139, 22 Fan Cooler Unit SW Valves Inservice test; March 8, 2011; 2-PT-M108, Residual Heat Removal, Safety Injection, and Containment Spray System Venting; March 10, 2011, 2-PT-Q17F, Alternate Safe Shutdown Supply Verification to 21 Safety Injection Pump / Residual Heat Removal Pump; and March 26, 2011, 0-SOP-Leakrate-001, RCS Leakrate Surveillance, Evaluation and Leak Identification.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of eight surveillance testing inspection samples as defined in NRC IP 71111.22.

b. Findings

No findings were identified.

1EP6 Drill Evaluation

EP Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of a routine Entergy emergency drill on February 3, 2011, to identify weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed emergency response operations in the simulator to determine whether the event classification, notifications, and protective action recommendations were performed in accordance with procedures. The inspectors also attended the station drill critique to compare inspector observations with those identified by Entergy staff in order to evaluate Entergys critique and to verify whether Entergy staff was properly identifying weaknesses and entering them into the CAP.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one sample as defined in NRC IP 71114.06.

b. Findings

No findings were identified.

2RS1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

Radiological Hazard Assessment The inspectors reviewed changes to plant operations that may result in a significant new radiological hazard for onsite workers or members of the public since the last inspection.

The inspectors reviewed whether Entergy personnel have assessed the potential impact of these changes and have implemented periodic monitoring, as appropriate, to detect and quantify the radiological hazard.

Recent radiological surveys from more than six plant areas were reviewed by the inspector to evaluate the thoroughness and frequency of the surveys and that they were appropriate based on the radiological hazards.

The inspectors conducted walkdowns and performed independent radiation surveys of the facility, including radioactive waste processing, storage, handling areas; and inside the Unit 3 containment, primary auxiliary building and spent fuel storage building, to evaluate the existing radiological conditions and the efficacy of the associated radiological postings and controls.

The inspectors observed and evaluated the following radiological risk-significant work activities:

Bullet nose repositioning on the Unit 3 reactor upper internals; Unit 3 reactor head shielding and established access controls; 32 reactor coolant pump seal replacement; Unit 3 reactor defueling activities; and Unit 3 spent fuel building fuel movement.

With respect to the above work activities, the inspectors reviewed whether appropriate pre-work surveys were performed and were sufficient to identify and quantify the radiological hazards and to establish adequate protective measures. In addition, the inspectors reviewed applicable radiological surveys associated with these work activities 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, the hazards associated with work activities that could negatively affect the radiological conditions, and any significant radiation field dose gradients that could result in non-uniform exposures of the body.

The inspectors selected at least five air sample survey records during refueling outage conditions and verified that the samples were collected and counted in accordance with licensee procedures. The inspectors observed work in potential airborne areas to evaluate if applicable air monitoring was representative of the breathing air zone of the workers. The inspector also reviewed the use of continuous air monitors (CAMs) to monitor real-time airborne conditions in accordance with Entergy procedures. The inspectors verified that Entergys program for monitoring loose surface contamination in areas of the plant was adequate to assess the potential for airborne contamination conditions.

Instructions to Workers The inspectors observed various radioactive material containers to verify that they were labeled and controlled in accordance with 10 CFR Part 20 requirements.

Radiation work permits (RWPs) associated with the radiological risk-significant work activities listed above, were evaluated by the inspectors to identify what work control instructions or control barriers were specified and that plant-specific TS high radiation area requirements were met, including the use of applicable electronic pocket dosimeter (EPD) alarm setpoints that were specified in conformance with survey indications and plant policy.

The inspectors reviewed one electronic personal dosimeter dose rate alarm occurrence that was documented in a CR. The inspectors verified that Entergy responded appropriately to the occurrence and that corrective actions and dose evaluations were adequate.

Contamination and Radioactive Material Control The inspectors conducted observations at the Unit 2 and Unit 3 radiological controlled area (RCA) egress locations to observe the performance of personnel surveying and releasing material for unrestricted use to verify that those activities were performed in accordance with plant procedures and the procedures were sufficient to control the spread of contamination and prevent unintended release of radioactive materials from the site.

The inspectors reviewed Entergys criteria for the survey and release of potentially contaminated material to verify that the radiation detection instrumentation was being used at its most effective sensitivity capability.

Radiological Hazards Control and Work Coverage During tours of the facility and review of the work activities listed above, the inspectors evaluated the ambient radiological conditions to verify that existing conditions were consistent with posted surveys, RWPs, and worker briefings, as applicable.

During these work activity performance observations, the inspectors reviewed whether the adequacy of radiological controls, such as required surveys (including system breach radiation, contamination, and airborne surveys), radiation protection job coverage (including audio and visual surveillance for remote job coverage), contamination controls, and the stations means of using EPDs in high noise areas as high radiation area (HRA) monitoring devices.

The inspectors reviewed whether radiation monitoring devices were placed on the individuals body appropriately to monitor dose from external radiation sources. This review included high-radiation work areas with significant dose rate gradients.

The inspectors reviewed five RWPs for work within potential airborne radioactivity areas with the potential for individual worker internal exposures. The inspectors evaluated the airborne radioactivity controls and monitoring, including potentials for significant airborne radioactivity levels (e.g., grinding, grit blasting, system breaches, entry into tanks, cubicles, reactor cavities). For these selected potential airborne radioactive areas, the inspectors reviewed the use of high-efficiency particulate air (HEPA) ventilation system operation.

The inspectors examined the licensees physical and programmatic controls for highly activated or contaminated materials (nonfuel) stored within the Unit 2 and Unit 3 spent fuel pools to verify that appropriate controls were in place to preclude inadvertent removal of these materials from the pool.

Tours within the RCA of Units 2 and 3 were conducted by the inspectors to evaluate radiological postings and physical controls for HRAs and very high radiation areas (VHRAs) with respect to regulatory requirements.

Risk-Significant High Radiation Area and Very High Radiation Area Controls The inspectors discussed with the Radiation Protection Manager and a first-line health physics supervisor, the controls and procedures for high-risk HRAs and VHRAs and actions to be taken during changing plant conditions.

Radiation Worker Performance During observation of the work activities listed above, the inspectors observed radiation worker performance with respect to applicable radiation protection work requirements to determine if workers were aware of the significant radiological conditions in their workplace and their work performance was within the RWP control/limit requirements specified for the work performed.

The inspectors reviewed several radiological problem reports since the last inspection that identified the cause of the event to be human performance errors to determine if there was an observable pattern traceable to a similar cause and if this perspective matched the corrective action approach taken by the licensee to resolve the reported problems.

Radiation Protection Technician Proficiency During observation of the work activities listed above, the inspectors evaluated the performance of radiation protection technicians with respect to radiation protection work requirements and determined if technicians were aware of the radiological conditions in their workplace and the RWP controls/limits and if their performance was consistent with their training and qualifications with respect to the radiological hazards and work activities.

The inspectors reviewed several radiological problem reports since the last inspection that identified the cause of the event to be radiation protection technician error to determine if there was an observable pattern traceable to a similar cause and if this perspective matched the corrective action approach taken by the licensee to resolve the reported problems.

Problem Identification and Resolution The inspectors reviewed whether problems associated with radiation monitoring and exposure control were being identified by Entergy personnel at an appropriate threshold and were properly addressed for resolution in the licensee CAP.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one sample as defined in NRC IP 71124.01.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

a. Inspection Scope

Inspection Planning

The inspectors reviewed the plant UFSAR to identify areas of the plant designed as potential airborne radiation areas and any associated ventilation systems or airborne monitoring instrumentation and a description of the respiratory protection program to include the location and quantity of respiratory protection devices stored for emergency use.

The inspectors reviewed the reported PIs to identify any related to unintended dose resulting from intakes of radioactive materials.

Engineering Controls During observation of the work activities listed in section 2RS1 of this report, the inspectors observed station personnels use of ventilation systems as part of its engineering controls (in lieu of respiratory protection devices) to control airborne radioactivity. In addition the inspectors reviewed the ventilation controls for the Unit 3 spent fuel storage building and Unit 3 containment during refueling conditions.

The inspectors evaluated several temporary HEPA ventilation systems used to support work in contaminated areas to verify that the use of these systems was consistent with Entergy procedural guidance and as low as is reasonably achievable (ALARA).

The inspectors observed the use of several CAMs within the RCA that were being used to monitor and warn personnel of changing airborne concentrations in the plant. The inspectors reviewed 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 ALARA.

Use of Respiratory Protection Devices During observation of the work activities listed in section 2RS1 of this report, the inspectors reviewed the use of respiratory protection devices and the use of engineering controls to limit the overall exposure of the workers. The inspectors verified that the respiratory protection devices used to limit the intake of radioactive materials were certified by the National Institute for Occupational Safety and Health/Mine Safety and Health Administration. The inspectors reviewed the respiratory protection qualification records of three respirator users to verify that these individuals were medically certified, fit tested and appropriately trained in the respirators that had been used. During work activity observations, the inspectors assessed the workers use of respiratory protection devices in the field.

The inspectors verified respiratory protection equipment storage and controls for the equipment staged and ready for use in the plant and stocked for issuance. The inspectors observed the physical condition of the equipment and applicable maintenance and inspection records for selected equipment that was ready for use.

Problem Identification and Resolution The inspectors reviewed problems associated with the control and mitigation of in-plant airborne radioactivity to verify issues were being identified by station personnel at an appropriate threshold, properly addressed for resolution in the CAP and that corrective actions were appropriate commensurate with the safety significance of the issues.

Specific documents reviewed during this inspection are listed in the attachment.

b. Findings

No findings were identified.

2RS4 Occupational Dose Assessment

b. Inspection Scope

Inspection Planning

The inspectors reviewed the results of radiation protection program audits related to internal and external dosimetry (i.e., licensees quality assurance (QA) audit).

The inspectors reviewed the most recent National Voluntary Laboratory Accreditation Program (NVLAP) report on the vendors most recent results to determine the status of the vendors external dosimetry program.

Entergys procedures associated with dosimetry operations and dose evaluations were reviewed to verify that Entergy has established procedural requirements for determining when external and internal dosimetry is required.

External Dosimetry NVLAP Accreditation The inspectors reviewed whether Entergys personnel dosimeters that require processing are NVLAP accredited. This review included the approved irradiation test categories for the type of personnel dosimeter used [optically stimulated luminescent (OSL)] that are consistent with the types and energies of the radiation present, and use of the dosimeters [e.g., to measure deep dose equivalent, shallow dose equivalent (SDE), and lens dose equivalent].

Passive Dosimeters (OSL)

The onsite storage of personnel dosimeters was evaluated by the inspectors to verify the appropriate background exposure monitoring of dosimeters was accounted for when not in use.

Active Dosimeters (Electronic Dosimeters)

The inspectors reviewed Entergy personnels use of a correction factor to address the response of the electronic dosimeter (ED) as compared to OSL for situations when the ED must be used to assign dose.

Internal Dosimetry Routine Bioassay (in vivo)

Entergy personnels use of passive monitoring using portal monitors for screening intakes was reviewed for adequacy to detect internally deposited radionuclides.

Positive whole body count records for 2010 were reviewed to verify that no detectable internal dose assessments were determined above 10 mrem.

Special Bioassay (in vitro)

During 2010, there were no internal dose assessments requiring in vitro monitoring for inspection review.

The inspectors reviewed and assessed the adequacy of Entergys program for dose assessments based on airborne/derived airborne concentration (DAC) monitoring. This review was to verify that flow rates and/or collection times for fixed head air samplers or lapel breathing zone air samplers were adequate to ensure that appropriate lower limits of detection are obtained. The inspectors reviewed the adequacy of procedural guidance used to assess dose when, if using respiratory protection, station personnel applies protection factors. There were no dose assessments that used airborne/DAC monitoring for 2010 to review.

Special Dosimetric Situations Declared Pregnant Workers The inspectors reviewed whether Entergy informs workers of the risks of radiation exposure to the embryo/fetus, the regulatory aspects of declaring a pregnancy, and the specific process to be used for (voluntarily) declaring a pregnancy.

There was one individual who declared their pregnancy during the current assessment period, and their exposure monitoring records and the stations program for limiting exposure for the declared pregnant worker were reviewed in accordance with the requirements of 10 CFR Part 20.

Dosimeter Placement and Assessment of Effective Dose Equivalent for External Exposures The inspectors reviewed the adequacy of Entergys methodology for monitoring external dose in situations in which non-uniform fields are expected or large dose gradients will exist (e.g., diving activities and SG jumps) to include criteria for determining dosimetry placement or the use of multiple badges.

Shallow Dose Equivalent

During 2010, there were no SDE dose assessments for inspection review.

Neutron Dose Assessment

The inspectors reviewed the stations neutron dosimetry program, including dosimeter type(s) and/or neutron survey instrumentation.

Problem Identification and Resolution The inspectors verified that problems associated with occupational dose assessment are being identified by Entergy personnel at an appropriate threshold and are properly addressed for resolution in the licensee CAP.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one sample as defined in NRC IP 71124.04.

b. Findings

No findings were identified.

2RS5 Radiation Monitoring Instrumentation

Inspection Planning

The inspectors reviewed the plant UFSAR to identify radiation instruments associated with monitoring plant radiological conditions including airborne radioactivity, process streams, effluents, materials/articles, workers, and post-accident monitoring, including those instruments used for emergency assessment.

The inspector reviewed a Quality Assurance audit that included Entergys onsite calibration facility.

The inspectors reviewed procedures specifying the methodology for performing instrument source checks and calibrations.

Walkdowns and Observations Walkdowns of five effluent radiation monitoring systems (including liquid and airborne monitoring) were performed by the inspector. The inspectors reviewed whether the material condition of the radiation monitoring systems to verify that effluent/process monitor configurations were aligned in accordance with offsite dose calculation manual (ODCM) and UFSAR descriptions.

The inspectors selected 10 portable survey instruments; 5 area radiation monitors (ARMs) and CAMs; and 4 personnel contamination monitors that were in use or available for issuance. Calibration records for the selected instruments were reviewed as well as currency of source checks, and instrument operability.

Calibration and Testing Program Process and Effluent Monitors The inspectors selected five effluent monitor instruments (including both liquid and gaseous monitors) and verified calibration and functional tests were performed consistent with radiological effluent TS/ODCM and that the licensee calibrates its monitors with a transfer standard instrument that is traceable to National Institute of Standards and Technology (NIST). In addition, the inspectors reviewed whether selected effluent monitor alarm setpoints were established as provided in the ODCM and station procedures.

Laboratory Instrumentation

The inspectors selected one of each type of laboratory analytical instrument used for radiological analyses (e.g., gross alpha, gross beta, proportional counters, gamma spectroscopy and liquid scintillation counters) to verify that daily performance checks and calibration data indicate that the frequency of the calibrations is adequate and there are no indications of degraded instrument performance.

Whole Body Counter Whole body counter calibration and functional check records were reviewed by the inspectors.

Post-accident Monitoring Instrumentation The inspectors selected the containment high-range monitors for both Units 2 and 3 and reviewed the calibration documentation since the last inspection for adequacy.

Contamination Monitors In-service personnel contamination monitors and small article monitors located in the Unit 2 and Unit 3 radiological controlled area egress point were selected to verify current calibration records and to verify that the alarm setpoint values are reasonable to ensure that licensed material is not released from the site.

Portable Survey Instruments, ARMs, Electronic Dosimetry, and Air Samplers/CAMs The inspectors reviewed calibration documentation for at least one of each type of instrument. For portable survey instruments and ARMs, the inspectors reviewed detector measurement geometry and calibration methods for each, which included the use of its instrument calibrators.

During review of calibration records of portable survey instruments, the inspectors screened astound calibration results and corresponding station technician actions for instruments found significantly out of calibration (greater than 50 percent).

Instrument Calibrator

The inspectors reviewed the basis for instrument calibrations and that the instrument calibrators used were calibrated using calibration transfer instruments traceable to NIST.

Calibration and Check Sources

The inspectors reviewed Entergys 10 CFR Part 61, Licensing Requirements for Land Disposal of Radioactive Waste, source term to determine if the calibration sources used were representative of the types and energies of radiation encountered in the plant.

Problem Identification and Resolution As documented in Section 4OA2 of this report, the inspectors verified that problems associated with radiation monitoring instrumentation are being identified Entergy personnel at an appropriate threshold and are properly addressed for resolution in the licensee CAP.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one sample as defined in NRC IP 71124.05.

b. Findings

No findings were identified

4OA1 Performance Indicator Verification

a. Inspection Scope

The inspectors sampled Entergy submittals for the below listed performance indicators (PIs) for the period from January 2010 through December 2010. To determine the accuracy of the PI data reported during those periods, the inspectors used definitions and guidance contained in Nuclear Energy Institute (NEI) Document 99-02, Regulatory Assessment Performance Indicator Guideline. As applicable, the inspectors reviewed operator narrative logs, issue reports, event reports, and NRC integrated inspection reports to validate the accuracy of the submittals. The inspectors also reviewed Entergys issue report database to determine if problems had been identified with the PI data collected or transmitted for these indicators.

Unplanned Scrams per 7000 Critical Hours (IE01)

Unplanned Power Changes per 7000 Critical Hours (IE03)

Unplanned Scrams with Complications (IE04)

Reactor Coolant System Activity (BI01)

Specific documents reviewed are described in the attachment to this report. These activities constitute completion of four PI samples as defined in NRC IP 71151.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review of Problem Identification and Resolution Activities

a. Inspection Scope

As required by IP 71152, Identification and Resolution of Problems, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that issues were being entered into Entergys CAP at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. 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 CAP. The inspectors reviewed attributes that included:

(1) complete and accurate identification of the problem;

(2) timely correction, commensurate with the safety significance;

(3) evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrences reviews; and

(4) classification, prioritization, focus, and timeliness of corrective actions.

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. Specific documents reviewed during this inspection are listed in the attachment.

b. Findings

No findings were identified.

.2 Annual Sample: Individual Rod Position Indication (RPI) System Problems

a. Inspection Scope

The inspectors performed an in-depth review of Entergy's evaluations and corrective actions associated with repetitive control RPI problems. Specifically, in 2006, Entergy personnel placed the RPI system in a Maintenance Rule (a)(1) status as a result of multiple failures of rod bottom bistables. Additional bistable failures occurred in 2007, 2008, and 2009. Failures of additional components, other than bistable, and repetitive instrument drift were also documented in various CRs by Entergy personnel.

The inspectors assessed Entergy's problem identification threshold, problem analysis, extent of condition reviews, compensatory actions, and the prioritization and timeliness of Entergy's corrective actions to determine whether Entergy personnel were appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken in accordance with the requirements of Entergy's CAP and 10 CFR 50 Appendix B. The inspectors performed field walkdowns, and interviewed plant operators and engineering personnel to assess the effectiveness of the implemented corrective actions, the reasonableness of the planned corrective actions, and to evaluate the extent of any on-going RPI problems. In addition, the inspectors reviewed operating procedures, operating logs, and interviewed licensed operators to assess operator response to RPI problems, including incorrect analog RPIs or loss of rod bottom lights, which occurred or might occur during reactor trip events. Specific documents reviewed are listed in the attachment to this report.

b. Findings and Observations

No findings were identified. Entergy personnel determined that the most probable cause of rod bottom bistable failures was age related degradation. In addition, Entergy personnel determined that electrolytic capacitors in the RPI circuitry were also susceptible to age related degradation, while the repetitive instrument drift was most probably related to analog circuitry design and calibration techniques. Entergy's corrective actions included a modification, completed in 2010, that replaced the original equipment bistables with new style bistables, revisions to instrument and control calibration procedures, and initiation of a corrective action to establish a PM task to periodically replace RPI electrolytic capacitors.

The inspectors determined Entergy's overall response to the issue was commensurate with the safety significance and included appropriate compensatory actions. The inspectors determined that the actions taken or planned were reasonable to resolve the bistable failure issue and improve RPI system performance. The inspectors identified a weakness in Entergy's evaluation of RPI problems because Entergy assumed the repetitive instrument drift was not due to any degraded material condition, but did not evaluate whether electrolytic capacitor degradation could also result in repetitive instrument drift. Entergy personnel entered the inspector's observations into their CAP as CR IP2-2011-00619.

Specific documents reviewed are described in the attachment to this report. These activities constitute completion of one identification and resolution of problems sample as defined in NRC IP 71152.

b. Findings

No findings were identified.

.3 Radiation Safety Cornerstone

a. Inspection Scope

(71124.01)

The inspectors reviewed one CR that was initiated between December 1, 2010 and January 10, 2011, CR IP2-2010-7316, that was associated with the radiation protection program. The inspectors reviewed whether problems identified by this CR was properly characterized in the stations event reporting system, and that applicable causes and corrective actions were identified commensurate with the safety significance of the radiological occurrence.

The inspectors reviewed eight corrective action CRs initiated between January and March 2011 that were associated with the radiation protection program. The inspector verified that problems identified by this CR were properly characterized in the licensees event reporting system, and that applicable causes and corrective actions were identified commensurate with the safety significance of the radiological occurrence.

b. Findings

No findings were identified.

4OA3 Event Follow-Up

Loss of the Normal Offsite Power Source

a. Inspection Scope

The inspectors reviewed the below listed event for plant status and mitigating actions to evaluate Entergy personnel performance and confirm that Entergy operators implemented actions and notifications (if required) in accordance with station procedures. The inspectors also reviewed Entergy's emergency response actions to evaluate Entergy staff performance and confirm that Entergy staff implemented actions and notifications in accordance with station procedures.

At 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br /> on March 1, 2011, Unit 2 experienced a loss of offsite power from the 138 kV circuit. The loss of the 138 kV circuit resulted in power loss to the station auxiliary transformer. The loss of the station auxiliary transformer de-energized 6.9 kV busses 5 and 6 and safety-related 480V buses 5A and 6A. All three EDGs automatically started as required and supplied power to buses 5A and 6A. No emergency action levels were entered because the second, independent offsite circuit, 13.8 kV, remained available throughout the event.

Operations personnel entered abnormal operating procedures 2-AOP-480V-1, Loss of Normal Power to any 480v Bus, and 2-AOP-138KV-1, Loss of Power to 6.9kv Bus 5 and/or 6, and transferred the power supply for buses 5A and 6A from EDGs to the safety-related 13.8kv offsite circuit. Entergy personnel performed an investigation and determined the loss of the 138kV circuit was the result of Con Edison work on feeder 95332 metering circuit test switch, when a small arc occurred causing an imbalance in the protective circuit which resulted in breakers isolating the 138 kV circuit. The operators terminated the event at 16:52 when the station auxiliary transformer was placed back in service and the normal 138 kV power was restored to all 480V buses.

Entergy personnel made a non-emergency 8-hr report to the NRC in accordance with 10 CFR 50.72(b)(3)(iv) for the actuation of the EDGs and will provide a written follow-up 60 day Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(iv).

The inspectors evaluated the response of control room personnel following the loss of the 138 kV circuit. The inspectors reviewed plant computer data, including evaluating plant data summary, plant parameter traces, personnel reports, and discussed the event with plant personnel to verify that plant equipment responded as expected and to ensure that operating procedures were appropriately implemented. The inspectors also reviewed whether Entergy's post trip review group (PTRG) identified the most probable cause(s) of the loss of the 138 kV circuit to facilitate corrective actions prior to swapping the 480V buses from the 13.8 kV to the 138 kV circuit. This event and the PTRG report were entered into Entergy's CAP as CR lP2-2011-1108.

The inspectors also reviewed station actions and decision making to verify decisions were consistent with a conservative approach to assessing the condition and in accordance with the site emergency plan. The inspectors reviewed logs and records from the event, interviewed operational and emergency planning staff, and reviewed corrective action documentation.

Specific documents reviewed during this inspection are listed in the attachment. These activities constitute completion of one event follow-up sample as defined in IP 71153.

b. Findings

No findings were identified associated with the operational response to the loss of the 138kV offsite circuit. The inspectors will conduct further review of the cause evaluation and associated corrective actions in conjunction with review of the Licensee Event Report to be submitted by Entergy personnel.

4OA5 Other Activities

.1 (Closed) Temporary Instruction 2515/179,Verification of Licensee Responses to NRC

Requirement for Inventories of Materials Tracked in the National Source Tracking System (NSTS) Pursuant to Title 10, Code of Federal Regulations, Part 20.2207 (10 CFR 20.2207)

a. Inspection Scope

The inspectors verified the information listed on Entergys inventory record by performing a physical inventory and visually identified each item listed on Entergys inventory. The inspectors verified the presence of the nationally tracked sources with an appropriate radiation survey instrument. During the physical inventory, the inspectors examined the physical condition of the shield devices containing nationally tracked sources, and evaluated the effectiveness of the licensees procedures for secure storage and handling of nationally tracked sources. The inspectors also verified that appropriate leak tests had been performed and determined that the posting and labeling of nationally tracked sources were adequate.

There had been no transfers or receipts of NSTS tracked sources from the licensees NSTS inventory since initial registration.

The inspectors reviewed the administrative information listed in the NSTS inventory for Indian Point Unit 2 and Unit 1 to ensure that the information was up to date. This information includes, but is not limited to:

• Mailing address;

• Physical or shipping address (for transmitting information via non-postal methods that cannot use a post office box);

• Telephone number, FAX number, and e-mail address for primary technical point of contact;

• Telephone number, fax number, and e-mail address for primary management point of contact; and

• The license numbers of NRC licenses that authorize the possession of nationally tracked source(s).

b. Findings

No findings were identified.

.2 Institute of Nuclear Power Operations (INPO) Plant Assessment Report Review

a. Inspection Scope

The inspectors reviewed the October 2009 final report for the INPO plant assessment of the Indian Point Nuclear Generating Station. The inspectors reviewed the report to ensure that issues identified were consistent with the NRCs perspectives of licensee performance and to identify significant safety issues that required further NRC follow-up.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On April 18, 2011, the inspectors presented an exit meeting of the inspection results of the integrated inspection to Mr. Joseph Pollock, Site Vice President, and other members of the Entergy staff. The licensee acknowledged the results of the inspection. No proprietary information was retained.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Entergy Personnel

J. Pollock Site Vice President

J. Abisamra Echelon Chief Engineer

R. Allen NDE Level III, Code Programs

H. Anderson Specialist - Nuclear Safety/License IV

N. Azevedo Supervisor - Engineering

J. Baker Shift Manager

S. Beagles Echelon Manager - Fleet Operations

M. Burney Specialist - Nuclear Safety/License IV

R. Burroni Manager - System Engineering

C. Childress Manager - Dry Cask Project

T. Cole Project Manager - NUC

G. Dahl Specialist - Nuclear Safety/License IV

R. Daley Engineer III - Nuclear

K. Davison Assistant Plant Manager

G. Dean Shift Manager

J. Dent Echelon General Manager - Plant Operations, Fleet Operations Support

D. Dewey Shift Manager

J. Dinelli Manager - Operations

B. Ford Echelon Senior Manager - Nuclear Safety and Licensing

T. Flynn Maintenance Inspection Coordinator

D. Gagnon Manager - Security

E. Harris Echelon Manager - Quality Assurance

G. Hocking Supervisor - Radiation Protection

F. Inzirillo Manager - IPEC Quality Assurance

D. Jacobs Echelon Senior Vice President - Planning, Development and Oversight

R. Lee Lead Engineer - Buried Pipe and Tank Program

J. Lijoi Superintendent - I&C

L. Lubrano Senior Lead Engineer

R. Mages Specialist - Senior HP/Chemical

T. McCaffrey Manager - Design Engineering

B. McCarthy Assistant Operations Manager

P. Morris Echelon Senior Staff Engineer

T. Motko System Engineer

T. Orlando Director - Engineering

T. Palmisano Echelon Vice President - Oversight

E. Primrose Shift Manager

S. Prussman Specialist - Nuclear Safety/License IV

J. Reynolds Specialist - Corrective Action

R. Robenstein Superintendent - Simulator

T. Salentino Superintendent - Dry Fuel Storage

S. Sandike Specialist - Senior HP/Chemical

P. Santini Senior Reactor Operator

A. Singer Superintendent - Licensed Operator Requalification Training

D. Smith Technical Specialist IV

T. Tankersly Echelon Director - Oversight

M. Tesoriero Manager - Programs and Components

A. Vitale General Manager - Plant Operations

R. Walpole Manager - Licensing

A. Williams Assistant General Manager - Plant Operations

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000247/2011002-01 NCV Main Steam System Configuration Control Procedure Not Adequate to Ensure Closure of MS-55D (Section 1R04)

05000247/2011002-03 NCV Entergy Personnel Did Not Identify a Leak on the Service Water Pump Piping (Section 1R19)

Opened

05000247/2011002-02 URI Notification Process for State/Local Authorities During a Simulator Scenario (Section 1R11)