IR 05000220/2010002

IR 05000220-10-002 and 05000410-10-002 on 01/01/10 - 03/31/10 for Nine Mile Point,

ML101260523
Person / Time
Site:	Nine Mile Point
Issue date:	05/06/2010
From:	Glenn Dentel Reactor Projects Branch 1
To:	Belcher S Nine Mile Point
References
IR-10-002
Download: ML101260523 (34)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION I

475 ALLENDALE ROAD KING OF PRUSSIA, PENNSYLVANIA 19406-1415 May 6, 2010 Mr. Sam Belcher Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 6~~

Lycoming, NY 13093 SUBJECT: NINE MILE POINT NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000220/2010002 AND 05000410/2010002

Dear Mr. Belcher:

On March 31, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Nine Mile Point Nuclear Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on April 16,2010, with you and 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 one NRC-identified and two self-revealing findings of very low safety significance (Green). Two of the findings were determined to involve violations of NRC reqUirements. However, because of the very low safety significance and because they are entered into your corrective action program (CAP), the NRC is treating these findings as non cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV noted in this report, 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 copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Senior Resident Inspector at Nine Mile Point Nuclear Station. In addition, if you disagree with the characterization of 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 I, and the NRC Senior Resident Inspector at Nine Mite Point Nuclear Station. The information you provide will be considered in accordance with Inspection Manual Chapter 0305. In accordance with 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 or from the Publicly Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html{the Public Electronic Reading Room).

Sincerely

.

L.fiJ;,d Glenn T. Dentel, Chief Projects Branch 1 DiviSion of Reactor Projects Docket Nos.: 50-220,50.410 License Nos.: DPR-63, NPF-69 Enclosure: Inspection Report 05000220/2010002 and 05000410/2010002 w/Attachment: Supplementallnformati,on cc w/encl: Distribution via UstServ

SUMMARY OF FINDINGS

IR 05000220/2010002,05000410/2010002; 01/01/2010 M03/31/2010; Nine Mile Point Nuclear

Station, Units 1 and 2: Maintenance Risk Assessment, Plant Modifications, and Followup of Events.

The report covered a three-month period of inspection by resident inspectors and announced inspections and an in-office inspection performed by regional inspectors. Three Green findings, two of which were non-cited violations (NCVs), were identified. The significance of most findings is Indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process (SOP)." The crossMcutting 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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREGM1649, "Reactor Oversight Process," Revision 4, dated December 2006.

Cornerstone: Initiating Events

Green.

A self-revealing finding of very low safety significance associated with a non-cited violation (NCV) of Technical SpeCification (TS) 5.4, "Procedures," was identified when technicians used an inadequate procedure for filling and venting a Unit 2 residual heat removal (RHR) system pressure detector following system maintenance, which resulted in an automatic scram. When the procedure was developed, Nine Mile Point Nuclear Station (NMPNS) did not identify that the detector to be filled and vented was connected to multiple detectors in other systems, and therefore did not evaluate the effect that the activity would have on these additional detectors. As immediate corrective action, RHR detector restoration was stopped and an investigation into the cause of the event was commenced.

The issLle was entered into the corrective action program (CAP) as condition report (CR)2010-0192.

The finding was more than minor because it was associated with the procedure quality attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additionally, the finding was similar to example 4.b in Appendix E of Inspection Manual Chapter {lMC) 0612, in that it resulted in a reactor scram. The finding was of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The finding had a cross-cutting aspect in the area of human performance, resources, because Nine Mile Point Nuclear Station (NMPNS) did not provide maintenance personnel with an accurate work package for filling and venting the 'C'

RHR prelssure detector (H.2.c per IMC 0310). (Section 40A3)

Cornerstone: Mitigating Systems

Green.

A self-revealing finding of very low safety significance was identified for inadequate coordination during concurrent execution of a maintenance procedure and an operating procedure, which resulted in a loss of power to the loads supplied by Unit 2 uninterruptible power supply (UPS) 2VBB-UPS1A. The loss of operational capabilities, and alarm and display functions, complicated normal plant operations and impacted an "anticipated transient without scram" (ATWS) mitigation strategy. As immediate corrective action, maintenance on UPS1A was stopped pending causal evaluation of the event. The issue was entE~red into the corrective action program (CAP) as condition report (CR) 2009-8928.

The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, the finding was significant because it would have impacted Nine Mile Point Nuclear Station's (NMPNS's)ability to execute emergency operating procedure N2-EOP-C5, "Failure to Scram," in that I the reactor manual control system was not available for use in accordance with N2-EOP-6,

Attachment 14, "Alternate Control Rod Insertions." The finding was of very low safety II*

significance because it was not a design or qualification deficiency, did not represent a loss of a system/train safety function, and did not screen as potentially risk significant due to ,.i external events. The finding had a cross-cutting aspect in the area of human performance, work control, because NMPNS did not address the impact of changes to the work activity on the plant and human performance (H.3.b per IMC 0310). (Section 1R13)

Green.

An NRC-identified finding of very low safety significance associated with a non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," was identified, in that Unit 2 Division 1 vital battery, 2BYS*BAT2A (the 2A battery), performance testing i was not performed in accordance with written test procedures. SpeCifically, procedure deviations were made during the three most recent biennial performance tests which resulted in inaccurate determinations of battery capacity. As immediate corrective action,

Nine Mile Point Nuclear Station (NMPNS) entered the issue into the corrective action program (CAP) as condition report (CR) 2010*1987 and implemented actions to estimate the current battery capacity. Based on the magnitude of the errors and current battery capacity margins, NMPNS determined that there were no operability issues with the 2A battery.

The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating levents to prevent undesirable consequences. Additionally, the finding was similar to example 2.c in Appendix E of Inspection Manual Chapter (IMC) 0612, in that the test performance issue was repetitive. The finding was of very low safety significance because it was not a design or qualification deficiency, did not represent a loss of a system/train safety function, and did not screen as potentially risk significant due to external events. The finding had a cross-cutting aspect in the area of human performance, work practices, because personnel did not follow the modified performance test procedure (H.4.b per IMe .

0310). (Section 1R18)other Findings None.

REPORT DETAILS

Summary of Plant Status

Nine Mile Point Unit 1 began the inspection period at full rated thermal power (R-rp). On January 18, power was reduced to 45 percent to secure reactorfeedwater pump (RFP) 13 for modification and testing of its flow control valve (FCV). Power was restored to full RTP on January 20. On February 16. power was reduced to 45 percent to secure RFP 13 for maintenance on its FCV. Power ascension to full RTP commenced on February 19, but was stopped at 90 percent due to a malfunction of tM RFP 13 FCV. Power was restored to full RTP the following day. On March 20, power was reduced to 72 percent for a control rod pattern exchange. Power was restored to full RTP later that day and remained there for the rest of the inspection period. During this period, Unit 1 also performed several brief power reductions to start or stop reactor recirculation pumps, to perform control rod pattern adjustments, and to conduct planned testing.

Nine Mile Point Unit 2 began the inspection period at full RTP. On January 7, an automatic scram occurred due to inadvertent actuation of the redundant reactivity control system (RRCS)that initiated alternate rod insertion (ARI) and recirculation pump trip (see Section 40A3). The scram was otherwise uncomplicated and a reactor startup was performed on January 9. Full RTP was achieved on January 11. Over the remainder of the inspection period, Unit 2 performed sE~veral brief power reductions to perform control rod pattern adjustments and, on March 28, entered end-of-cycle coast down (gradual power reduction due to fuel depletion). At the end of the inspection period, Unit 2 was operating at approximately 99 percent RTP.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems. and Barrier Integrity

1R04 Eguipment Alignment

.1 Partial System Walkdown (71111.04Q - Five samples)

a. Inspection Scope

The inspectors performed partial system walkdowns to verify risk~significant systems were properly aligned for operation. The inspectors verified the operability and alignrnent of these risk-significant systems while their redundant trains or systems were inoperable or out of service for maintenance. The inspectors compared system lineups to system operating procedures, system drawings, and the applicable chapters in the updated final safety analysiS report (UFSAR). The inspectors verified the operability of critical system components by observing component material condition during the system walkdown.

The following plant system alignments were reviewed:

• Unit 1 containment spray raw water (CSRW) system 11 (111 and 112) due to increased risk significance while CSRW system 121 was inoperable and unavailable during pump rebuild; I

• Unit 'I liquid poison system 11 due to increased risk significance during liquid poison pump 12 overhaul;

• Unit 2 'A' residual heat removal (RHR) system due to increased risk significance I while the 'B' and 'C' RHR systems were inoperable during maintenance; I"

• Unit 2 low pressure core spray (LPCS) system due to increased risk significance

• while the 'c' RHR system was inoperable during maintenance; and Unit 2 high pressure core spray (HPCS) system due to increased risk significance I

while the reactor core isolation cooling (RCIC) system was inoperable during maintenance.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Routine Resident Inspector Tours (71111.050 - Six samples)

a. Inspection Scope

The inspectors toured areas important to reactor safety to evaluate the station's control of transient combustibles and ignition sources, and to examine the material condition.

operational status. and operational lineup of fire protection systems including detection, suppression. and fire barriers. The inspectors evaluated fire protection attributes using the criteria contained in Unit 1 UFSAR Appendix 10A. "Fire Hazards Analysis," and Unit 2 procedure N2-FPI-PFP-0201, "Unit 2 Pre-Fire Plans." The areas inspected included:

• Unit 1 reactor building (RB) 237 foot elevation;

• Unit 1 cable spreading room, turbine building (TB) 250 foot elevation;

• Unit 1 TB 261 foot elevation;

• Unit 2 RB 175 foot elevation;

• Unit 2 DiviSion 2 switchgear room, control building 261 foot elevation; and

• Uni12 service water (SW) pump rooms, screenwell building 224 and 261 foot e~levations.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance

.1 AnnlJal Heat Sink Performance (71111.07A - One sample)

a. Inspection Scope

The inspectors observed the performance of the Unit 2 'AI RHR heat exchanger performance testing in accordance with N2-TTP-RHS-4Y003, "Residual Heat Removal System Heat Exchanger (2RHS*E1A) Performance Monitoring (Suppression Pool Cooling Mode)," Revision 01. The inspectors reviewed the test results to verify that system performance was consistent with the design basis as specified in the UFSAR.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Regualification Program

.1 Quarterly Review (71111.11 Q - Two samples)

a.

InsRection Scoge The inspectors evaluated two simulator scenarios in the licensed operator requalification training (LORT) program. The inspectors assessed the clarity and effectiveness of communications, the implementation of appropriate actions in response to alarms, the perfclrmance of timely control board operation, and the oversight and direction provided by the shift manager. During the scenario, the inspectors also compared simulator performance with actual plant performance in the control room. The following scenarios were observed:

• On February 25, 2010, the inspectors observed Unit 1 LORT to assess operator and instructor performance during a scenario involving loss of power board 16B. a stuck open electromatic relief valve, a steam leak in the drywell, and a manual reactor pressure vessel blowdown. The inspectors evaluated the performance of risk significant operator actions including the use of special operating procedures (SOPs) and emergency operating procedures (EOPs).

• On March 2, 2010, the inspectors observed Unit 2 LORT to assess operator and instructor performance during a scenario involving failure of the automatic reactor vessel water level control system, loss of offsite electrical power line 6 with failure of the Division 2 emergency diesel generator (EDG) to automatically start, loss of the re~maining offsite electrical power line which led to an automatic reactor scram, opening and subsequent failure to close of a reactor safety relief valve with a leak in the downstream tailpiece that resulted in a steam leak in the drywell, failure of the 'A' RHR pump due to a motor electrical fault, and failure of the 'B' RHR containment spray valves to open. The inspectors evaluated the performance of risk significant operator actions including the use of SOPs and EOPs.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12Q - One sample)

a. Inspection Scope

The inspectors reviewed performance-based problems, and the performance and condition history for the Unit 1 feedwater system, to assess the effectiveness of the maintenance program. The inspectors reviewed the systems to ensure that the station's review focused on proper maintenance rule scoping in accordance with 10 CFR Part 50.65, characterization of reliability issues, tracking system and component unavailability, and 10 CFR Part 50.65(a)(1} and (a}(2) classification. In addition. the inspectors reviewed the site's ability to identify and address common cause failures, and to trend key parameters.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - Six samples)

a. Inspection Scope

The inspectors evaluated the effectiveness of the maintenance risk assessments required by 10 CFR Part 50.65(a)(4). The inspectors reviewed equipment logs, work schedules, and perfonned plant tours to verify that actual plant configuration matched the assessed con'figuration. Additionally, the inspectors verified that risk management actions for both planned and emergent work were consistent with those described in station procedures. The inspectors reviewed risk assessments for the activities listed below,

• Week of January 18, that included a power reduction to 45 percent to secure RFP 13 for modification and testing of its flow control valve, SW pump 11 breaker maintenance, emergency service water pump 12 breaker maintenance, and emergent maintenance to identify the cause of an automatic swap of the RFP 13 flow control valve controllers and to remove the SW pump 11 motor for off-site rebuild after low winding resistance readings were identified.

• Week of February 8, that included containment spray raw water pump 121 rebuild.

reactor building closed loop cooling system heat exchanger 12 cleaning, contro~

room chiller 12 maintenance, and an emergent four day maintenance period for EDG 103 to troubleshoot its failure to start during a monthly surveillance.

• Week of March 8, that included channel 12 reactor recirculation flow converter calibrations, corrective maintenance on channel 11 of the RFP 13 flow control valve control system, corrective maintenance on instrument air compressor 12, a power reduction to 94 percent for turbine valve testing, and emergent maintenance to replace the charcoal filter media in both trains of the reactor building emergency ventilation system (RBEVS) due to the 11 train having failed its surveillance for iodine removal efficiency.

• Week of December 28, that included SW pump 'D' discharge strainer maintenance, installation of an inlet drain valve for instrument air dryer 3B, and preventive maintenance on non-vital uninterruptible power supply (UPS) 2VBB-UPS1A.

• Week of February 8, that included a two day maintenance period for the RCIC system, RCIC system quarterly surveillance, 'A' RHR heat exchanger performance testing, maintenance on non-vital UPS 2VBB-UPS1 B (a loss of the associated loads would result in a scram with complications). maintenance on the off-site power supply breaker to non-vital switchgear 2NPS-SWG001 that would preclude its automatic fast transfer to off-site power if required, and replacement of two jumpered cells in the Division 1 vital battery.

• Week of February 22, that included Division 3 EDG monthly surveillance and cycle surveillance to test the load shedding circuit, a three day maintenance period for Division 3 switchgear that caused the HPCS system to be inoperable, modification of splices in the main power cables to the HPCS pump, Division 2 standby liquid control system quarterly surveillance, a power reduction to 70 percent for a control rod line adjustment, and emergent operations to restore the reactor water cleanup system to service following shutdown due to failure of a pump seal and to troubleshoot and repair intermittent locking up of the 'B' RFP flow control valve.

b. Findings

Introduction.

A self-revealing finding of very low safety significance (Green) was Identified on December 31,2009, for inadequate coordination during concurrent execution of a maintenance procedure and an operating procedure, which resulted in a loss of power to the loads supplied by Unit 2 UPS 2VBB-UPS1A. The loss of operational capabilities, and alarm and display functions, complicated normal plant operations and impacted an "anticipated transient without scram" (ATWS) mitigation strategy.

DesGription. During the week of December 28,2009, electrical maintenance personnel were performing scheduled maintenance on 2VBB-UPS1A (UPS1A) in accordance with electrical preventative maintenance procedure N2-EPM-GEN-V624, "UPS Inverter Functional Checks, Cleaning and Inspection," Revision 00900. This procedure is normally performed over a five day period, and sections in the procedure are sequenced to le:3ve the eqUipment in the condition required to commence work in the following section. However, on this occasion, the steps were being performed out of sequence in I

an attempt to streamline the process, such that the procedure could be completed in four days. This was considered to be acceptable because the procedure contains an allowance for sections of the procedure to be performed out of sequence, relying on the I

experience of the maintenance personnel to establish the required equipment initial conditions.

Prior to commencing maintenance on December 31.2009, UPS1A was energized and carrying loads. The initial conditions for the portion of the procedure to be performed were that the UPS was energized and carrying no loads, in preparation for UPS shutdown. The electrical maintenance personnel contacted Operations personnel in the control room to discuss how to proceed. It was agreed that maintenance personnel would perform steps to shut down UPS1A per operating procedure N2-0P-71 D.

"Uninterruptible Power Supplies (UPS)," Revision 00500. in coordination with steps in N2-EPM-GEN-V624, to establish the desired maintenance conditions. This plan was not adequately coordinated during implementation, and resulted in UPS1A being shut down before its loads had been transferred to the maintenance power supply.

Although UPS1A is not a safety class UPS. it does supply a number of systems and indications that are important to plant operation. Among the loads that are lost upon a loss of UPS1A are: the reactor manual control system, which allows operators to manually insert or withdraw control rods; the reactor full core display, which provides status indicating lights and alarms for all control rods; the hydraulic power units for the

'A' reactor recirculation flow control valve, which allows manual adjustment of '

recirculation flow for control of reactor power; and numerous digital displays in the control room. Control room operators responded in accordance with special operating proc(~dure N2-S0P-71, "Loss of 2VBB-UPS1A, 1B. 1G," Revision 00500, and implElmented compensatory actions where required. Plant conditions had been stable prior to the loss of UPS1A, which simplified the required response. After the cause of the problem was fully understood, operators reenergized UPS1A and restored power to its loads within approximately three hours.

As immediate corrective action, maintenance on UPS1A was stopped pending causal evaluation of the event. The issue was entered into the corrective action program (CAP)as condition report (CR) 2009~8928. NMPNS is evaluating long term corrective action to revise N2~EPM-GEN-V624 to prevent non-sequential performance of sections of the procedure.

Analysis.

The inspectors determined that the loss of UPS1A loads due to inadequacies with the procedure for deenergizing the UPS to support maintenance was a performance deficiency. Constellation Fleet Program Directive CNG-MN-1.01, "Conduct of Maintenance," Revision 0000, requires that, when maintenance cannot be accomplished as described by the approved procedure, the work shall be stopped and shall not resume until the procedure is changed to reflect the correct work practice. N2 EPM-GEN-V624 inappropriately relied on craft experience to establish conditions that should have been prescribed by written procedure. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability. reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, the finding was significant because it would have impacted NMPNS's ability to execute emergency operating procedure N2 EOP-C5, "Failure to Scram," in that the reactor manual control system was not available for use in accordance with N2*EOP-6, Attachment 14. "Alternate Control Rod Insertions." The inspectors evaluated the significance of this finding using Inspection Manual Chapter (IMC) 0609, Attachment 4, "Phase 1 -Initial Screening and Characterization of Findings." The finding was of very low safety significance because it was not a design or qualification deficiency, did not represent a loss of a*system/train safety function, and did not screen as potentially risk significant due to external events.

The finding had a cross.cutting aspect in the area of human performance, work control, because NMPNS did not address the impact of changes to the work activity on the plant and human performance (H.3.b per IMC 0310).

Enforcement.

No violation of regulatory requirements occurred. The inspectors determined that the finding did not represent a noncompliance issue because UPS1A is not a safety class component and therefore is not subject to the requirements of 10 CFR .

Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants." (FIN 0500041012010002-01, Inadequate Maintenance Procedure Results in Loss of Loads for Non-Vital UPS)

1R15 Operability Evaluations (71111.15 - Six samples)

a. Inspection Scope

The inspectors evaluated the acceptability of operability evaluations, the use and control of compensatory measures. and compliance with technical specifications (TS). The evaluations were reviewed using criteria specified in NRC Regulatory Issue Summary 2005-20, "Revision to Guidance Formerly Contained in NRC Generic Letter (GL) 91-18,

'Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability'," and Inspection Manual Part 9900, "Operability Determinations and Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety. The inspectors' reviews included yerification that the operability It determinations were made as specified by Procedure CNG-OP-1.01-1002, "Conduct of Operability Determinations I Functionality Assessments." The technical adequacy of the determinations was reviewed and compared to the TSs, UFSAR. and associated design basis documents (DBDs). The following evaluations were reviewed:

• CR 2010-0697, Electrical Design Engineering Calculation EC-042, Revision 11, concerning the acceptability of two replacement cells in the Unit 2 Division 1 vital battery 2BYS*BAT2A with respect to TS-required biannual surveillance testing;

• CR 2010-1246 concerning operability of Unit 1 EDG 103 following troubleshooting activities that were unsuccessful at definitively identifying the cause of a previous failure to start;

• GRs 2010-1494 and 2010-1495 concerning the effect of minor sulfation (deposition clf lead sulfate), that is occurring on the negative plates of Unit 1 safety class batteries 11 and 12, on current operability of the batteries, and the potential for future performance degradation;

• CR 2010-1597 concerning the affect of incorrectly sized wiring and splices associated with solenoid operated valves on the functionality of Unit 1 instrument air compressor 12;

• CR 2010-1669 concerning the affect of anomalous indications of reactor pressure on operability of the Unit 2 RRCS; and

• CR 2010-2106 concerning the implications of RBEVS 11 charcoal filter exhaustion for RBEVS 12.

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

.1 Temoorary Modifications (One sample)

a. Inspection Scope

The inspectors reviewed a Unit 2 temporary plant modification to jumper two cells in the Division 1 vital battery. The modification was implemented through Engineering Change Packages (ECPs) ECP-10-000097 and ECP-10-000115, "Develop a Temporary Change to Jumper Cell 11

(20) on Battery 2BYS*BAT2A." The change was necessary because the subject cells had developed cracks in their casings, Which could degrade the battery's capability if the cracks were to spread such that a significant amount of electrolyte could leak out. The inspectors reviewed the 10 CFR Part 50.59 screening against the system design bases documentation to verify that the modification did not affect system operability. The inspectors verified the adequacy of acceptance testing and performed a walkdown of the installed modification.

b. Findings

Introduction:

The inspectors identified a finding of very low safety significance (Green)aSSOCiated with a non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," in that, Unit 2 Division 1 vital battery performance testing was not performed in accordance with written test procedures. Specifically, procedure deviations were made during the three most recent biennial performance tests which resulted in inaccurate determinations of battery capacity.

Description:

ECP-1 0-000115 was written to demonstrate that the Unit 2 Division 1 battery, 2BYS*BAT2A (the 2A battery). retained adequate capacity for operability with two cells electrically bypassed Gumpered out). The inspectors reviewed the change package, along with recent test procedures and test results for the 2A battery, to verify the design input to the evaluation.

Performance (capacity) tests are TS-required tests that are performed for safety related batteries. Industry guidance for battery performance testing is contained in Institute of Electrical and Electronics Engineers (IEEE) 450 [series], "Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary

. Enclosure Applications." During a performance test, the battery is discharged at a vendor specified rate until the battery is fully discharged (105 volts direct current (Vdc), or an average of 1.75 Vdc per cell). NMPNS performs modified performance tests at Unit 2, whiGh fulfill the requirements of a performance test and a service test (a separate TS required discharge test) by having a higher discharge rate for the first minute of the test.

For a performance test or modified performance test, the battery capacity is calculated basl~d on the amount of time until the battery becomes fully discharged. The calculated capacity, when trended and properly evaluated, will accurately determine when a battery is re'aching the end of its service life. Accurate measuring and trending of the battery's capacity are required to establish the correct testing frequency. For the 2A battery, the capacity was also used as a basis to reduce the battery aging margin in ECP-10 000115.

During a performance or modified performance test, it is possible that a cell will approach reversal (a condition in which the cell acts as a load on the rest of the battery),which is identified by the cell reaching an indiVidual cell voltage of 1.0 Vdc. NMPNS's modified performance test procedure, N2-ESP-BYS-R685, "Div 11111111 Battery Modified Profile Test." Revisions 2 and 3, states, "Any cell voltage that falls to 1.0 Vdc will result in shutdown of test. Test shall be considered a failed test." To anticipate this potential, 1\l2-ESP-BYS-R685 requires that the computer test equipment be set up such that the test will automatically stop if any cell reaches 1.0 Vdc.

The inspectors reviewed the three most recent modified performance tests from 2004, 2006, and 2008. The inspectors identified that these tests were not performed in accordance with the station's written procedure or with the industry guidance contained in IEEE 450.

In 2004, cell 60 reached 1.0 Vdc when the test was nearing completion. Contrary to the prooedure, the test was stopped when the cell reached 1.0 Vdc, and the test was documented as successful. Since cell 60 reached 1.0 Vdc prior to the battery reaching its minimum voltage, the battery capacity was recorded as being lower than it actually was, which impacts accurate trending of capacity.

In 2006, cell 60 again reached 1.0 Vdc when the test was nearing completion. As designed, the test equipment stopped the test automatically. However, the computer test data indicates that maintenance personnel resumed the test after a brief pause.

This action was contrary to the procedure. Pausing the test at this critical time allowed the battery to inappropriately recover, which resulted in the capacity being recorded as higher than it actually was. This test was also documented as successful.

In 2008, the test equipment was not set up in accordance with the procedure. This resulted in the test ending after the first minute. Maintenance personnel resumed the test in manual control, without procedural guidance for doing so. Because of the difficulty with maintaining a load bank in manual, the load did briefly drop below the minimum value. Although there is a step in the procedure to verify that the minimum load is reached and to record the value, no mention was made of the excursion below the minimum value. As the test neared completion, cell 60 reached 1.0 Vdc. The computer test data shows that the test equipment stopped the test as expected, but that maintenance personnel manually resumed the test. .As with the 2006 test, pausing the test at this point allowed the battery to inappropriately recover, which resulted in recording a capacity that was higher than actual. This test was documented as successful.

The inspectors concluded that NMPNS failed to perform TS-required testing for battery 2BYS*BAT2A in accordance with written test procedures for the years 2004 through 2008. The result was that the battery capacity had not been accurately measured for three test cycles. Without accurate battery capacity measurements, the TS-required values and testing frequencies are not assured.

NMPNS entered the issue into the CAP as CR-2010-1987 and implemented actions to estimate the current battery capacity. Based on the magnitude of the errors and current battery capacity margins, NMPNS determined that there were no operability issues with the 2A battery. The inspectors independently evaluated the battery capacity and similarly concluded that the issues identified did not render the 2A battery inoperable with two cells jumpered out. The inspectors therefore concluded that ECP-10-000115 was acceptable.

Analysis:

The inspectors determined that the failure to properly perform battery testing in accordance with written test procedures was a performance deficiency. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, the finding was similar to example 2.c in Appendix E of IMC 0612. in that the test performance issue was repetitive. The inspectors evaluated the significance of this finding using IMC 0609.

4. "Phase 1 -Initial Screening and Characterization of Findings." The

finding was of very low safety significance because it was not a design or qualification defiCiency, did not represent a loss of a system/train safety function, and did not screen as potentially risk significant due to external events. The finding had a cross-cutting aspe;ct in the area of human performance, work practices, because personnel did not follow the modified performance test procedure (HA.b per IMC 0310).

Enforcement:

10 CFR Part 50, Appendix B, Criterion XI, "Test Control," states, in part, that, "A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components win perform satisfactorily in service is ... performed in accordance with written test procedures ..." Contrary to the above, on three occasions from March 22, 2004, to March 24,2008, TS-required performance testing for safety related battery 2BYS*BAT2A was not completed in accordance with Procedure N2-ESP-BYS-R685, "Div 11111111 Battery Modified Profile Test," Revisions 2 and 3, as applicable. Specifically, deviations were made from the battery performance tests which resulted in inaccurate determinations of battery capacity. Because this violation was of very low safety significance and has been entered into the CAP as CR 2010-1987, this violation is being treated as an NeV, consistent with the NRC Enforcement Policy. (NCV 0500041012010002*02, Inadequate Performance Testing of the Division 1 Battery)

.2 Permanent Modifications (One sample)

a. Inspection Scope

The inspectors reviewed one Unit 1 permanent plant modification, ECP-09~000515, "Quick Trak Firmware Upgrade." The purpose of this change was to ensure proper operation of the transfer function of the turbine-driven feedwater pump flow control valve positioner system. The inspectors reviewed the associated 10 CFR Part 50.59 screening against system design basis information, verified that postwinstaHation tests were adequate, and that NMPNS controlled the modification in accordance with station procedures.

b. Findings

No findings of significance were identified.

1 R19 PostwMaintenance Testing (71111.19 - Six samples)a.

InsRE~ction Scope The inspectors reviewed the post maintenance tests (PMTs) listed below to verify that procedures and test activities ensured system operability and functional capability. The inspectors reviewed the test procedure to verify that the procedure adequately tested the safety functions that may have been affected by the maintenance activity, that the acceptance criteria in the procedure were consistent with information in the applicable licensing basis and/or DBDs, and that the procedure had been properly reviewed and approved. The inspectors also witnessed the test or reviewed test data, to verify that the tE1St results adequately demonstrated restoration of the affected safety functions.

• Unit 1, Work Order (WO) C90721202 to install a firmware modification in the Unit 1 RFP 13 FCV control system. The PMT was to verify proper operation of the valve actuator, the ability to transition between controllers, and operation of position indication and alarms, in accordance with N1-MFT-104. "Modification Test for Mod N1-06-023, FCV~29-134 Actuator," Revision 00101.

• Unit 1. WO C90684952 to overhaul liquid poison pump 12. The PMT was to verify proper pump operation in accordance with N1-ST-Q8B, "Liquid Poison Pump 12 and Check Valve Operability Test," Revision 0000.

• Unit 1, WO C90808077 to replace the charcoal filter in RBEVS 12. The PMT was to verify proper system operation in accordance with N1-TSP-202-001, "Testing of Unit 1 Reactor Building Emergency Ventilation System," Revision 02.

• Unit 1, WO C90605552 to overhaul containment spray raw water pump 121. The PMT was to verify proper pump operation in accordance with N1-ST-Q6B, "Containment Spray System Loop 121 Quarterly Operability Test," Revision 00900.

• Unit 2, WO C90657598 to overhaul the actuator for the 'B' RHR pump minimum flow valve, 2RHS*MOV4B. The PMT was a diagnostic test in accordance with S-EPM GEN-064, "Acquisition, Analysis, and Trending of MC2 Data," Revision 00400.

• Unit 2, WO C90629050 to applynew environmentally qualified (EQ) insulation to the HPCS pump motor power lead connections. The PMT was to measure insulation resistance to ground and to run the pump in accordance with N2-0SP-CSH-Q@002.

"HPCS Pump and Valve Operability and System Integrity Test," Revision 00400.

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities (71111.20 - One sample)

a.

InspE~ction Scope The inspectors observed and reviewed the following activities during the Unit 2 forced outane from January 7 through January 9,2010.

The inspectors observed portions of the plant shutdown and verified that the TS requirements with respect to reactor coolant system (RCS) cooldown limitations were satisfied. The inspectors reviewed outage schedules and procedures, and verified that TS-specified safety system availability was maintained and that shutdown risk was considered.

The inspectors observed portions of the reactor startup following the outage, and verified through control room observations, discussions with personnel, and log reviews that safety-related equipment specified for mode change was operable.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22 - Seven samples)

a. Inspection Scope

The inspectors witnessed performance of and/or reviewed test data for risk-significant surveillance tests (STs) to assess whether the components and systems tested satisfied design and licensing basis requirements. The inspectors verified that test acceptance criteria were clear, demonstrated operational readiness and were consistent with the DBDs; that test instrumentation had current calibrations and the range and accuracy for the application; and that tests were performed, as written, with applicable prerequisites satisfied. Upon test completion, the inspectors verified that equipment was returned to the status specified to perform its safety function.

The following STs were reviewed:

• N1-ST-Q1B, "CS [core spray] 121 Pump, Valve and SDC Water Seal Check Valve Operability Test," Revision 0100;

• N1-ST-Q6C, "Containment Spray System 112 Quarterly Operability Test," Revision 00801 ;

• N1-ST-M4B, "Emergency Diesel Generator 103 and PB [power board] 103 Operability Test," Revision 02;

• N2-0SP-CSH-Q@002, "HPCS Pump and Valve Operability and System Integrity Test," Revision 00400;

• N2-0SP-RHS-Q@006, "RHR System Loop C Pump and Valve Operability Test and System Integrity Test," Revision 00100;

• N2-0SP-ICS-Q@002, "RCIC Pump and Valve Operability Test and System Integrity Test and ASME XI Functional Test," Revision 07; and

• U2, N2-ISP-ADS-Q003, "Quarterly Functional Test and Calibration of the ADS

[automatic depressurization system] Logic," Revision 00300 .

. This represented a total of seven inspection samples, of which two were Routine Surveillances and five were In-Service Testing as defined by Inspection Procedure 71111.22.

No findings of significance were identified.

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation (71114.06 - One sample)

a. Inspection Scope

The inspectors observed simulator and technical support center activities associated with the Unit 2 emergency planning drill on February 16, 2010. The scenario consisted of a fire in the only operable unit cooler in the Division 2 SW pump room, failure of Division 1 vital switchgear 2EJS*US1 that resulted in a Joss of all Division 1 loads, a react()r coolant leak in the drywell that led the operators to insert a manual scram due to high drywell pressure, failure of RCIC due to high turbine exhaust pressure, failure of the HPCS injection valve to open, primary containment failure, and high radiation levels in the secondary containment that led operators to perform a reactor pressure vessel blowdown. The inspectors evaluated the emergency classification declarations and notifications against the requirements contained in 10 CFR Part 50.72. 10 CFR Part 50, Appendix E, and the Nine Mile Point emergency plan implementing procedures.

b. Findings

No findings of significance were identified.

I I

OTHER ACTIVITIES

40A1 Performance Indicator Verification (71151 - Six samples)

a. Inspection Scope

The inspectors sampled NMPNS submittals for the performance indicators (Pis) listed below. The PI definition guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6, was used to verify the basis in reporting for each data element and the accuracy of the PI data reported.

Cornerstone: Initiating Events

The inspectors reviewed licensee event reports and operator logs to determine whether NMPNS accurately reported the number of unplanned scrams, unplanned scrams with complications, and unplanned power changes at Unit 1 and Unit 2 from January 2009 through December 2009.

b. Findings

No findings of significance were identified.

40A2 Problem Identification and Resolution (71152)

.1 ReviEfw of Items Entered into the CAP

a. Inspection Scope

As specified by Inspection Procedure 71152. "Identification and Resolution of Problems," and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into NMPNS's CAP. In accordance with the baseline inspection procedures, the inspectors also identified selected CAP items across the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for additional follow-up and review. The inspectors assessed the threshold for problem identification, the adequacy of the cause analyses, extent of condition review, operability determinations, and the timeliness of the specified corrective actions.

b.

Findin.fl§ No findings of significance were identified.

40A3 Followup of Events and Notices of Enforcement Discretion (71153)

.1 Unit 2 Reactor Scram Due to Inadvertent Actuation of the Redundant Reactivity Control

System a.

InsRection Scope At 1:00 a.m. on January 7, 2010, Unit 2 scrammed from full RTP due to inadvertent actuation of the RRCS while filling and venting a pressure detector in the 'C' RHR system. The RRCS actuation produced an ARI scram and caused the reactor recirculation pumps to shut down due to the recirculation pump trip (RPT) function. The detector fill and vent also caused RCIC to initiate due to an invalid reactor vessel low water level indication.

Operators responded in accordance with the applicable emergency operating procedures. RCIC automatically shut down as designed when reactor vessel water level reached the high level set point following the scram. Operators then continued to raise water level as required to establish natural circulation cooling. Due to the injection of relatively cold water by the control rod drive system following the scram arid the lack of operating reactor recircUlation pumps, the reactor vessel cooldown rate limit for a one hour period was slightly exceeded; the TS-required corrective actions for this condition were completed prior to the subsequent reactor startup. The scram was otherwise uncomplicated.

The inspectors responded to the control room and observed operators' responses to the event. The inspectors verified that operators responded in accordance with the applicable procedures. The inspectors confirmed that no emergency plan emergency action level thresholds had been exceeded and that the event was appropriately reported to the NRC.

The inspectors reviewed the circumstances surrounding the event. The inspectors monitored startup preparation activities and corrective actions through attendance at outage update meetings, discussions with plant personnel, and review of records, including the post-scram review.

b.

Findir19§.

Introduction.

A self-revealing finding of very low safety significance (Green) associated with 81 NCV of TS 5.4, "Procedures," was identified on January 7,2010, when technicians used an inadequate procedure for filling and venting a Unit 2 RHR pressure detector following system maintenance, which resulted in an automatic scram. When the procedure was developed, NMPNS did not identify that the detector to be filled and vented was connected to multiple detectors in other systems, and therefore did not evaluate the effect that the activity would have on these additional detectors.

Description.

During the week of January 4, 2010, Unit 2 personnel were performing several maintenance activities on the 'B' and 'c' RHR systems that required a portion of the systems to be drained. On January 6, 2010, technicians were refilling 'C' RHR pressure detectors as part of the maintenance restoration. While preparing to fill and vent differential pressure detector 2RHS"'PDT24C, they observed that the pressure sensing line was also connected to a detector in the 'B' RHR system, 2RHS"'PDT24B, and that this detector had not been removed from service as part of the maintenance activity. They contacted the control room to obtain further guidance, since this configuration was not indicated in the work order. The control room senior reactor oper;ator (SRO) reasoned that it would be acceptable for the technicians to proceed, despite the unexpected configuration, because both the 'B' and 'C' RHR systems were inopE~rable. As the technicians proceeded to fill and vent 2RHS"'PDT24C, a reactor scram occurred at 1:00 a.m., January 7,2010, due to actuation of the ARI function of the HRCS.

The RRCS actuates on either high reactor pressure or low reactor water level. Review of th~:l detector filling activity revealed that 2RHS"'PDT24C is connected to multiple other detectors besides 2RHS"'PDT24B through a common reference Hne, and that one of thesE~ detectors (2ISC"'PT4B) is for the low reactor vessel water level input to the RRCS.

NMPNS concluded that operation of 2RHS*PDT24C had perturbed the RRCS detector such that it sensed an invalid transient low reactor vessel water level cond.ition, and thereby initiated the system. As immediate corrective action, RHR detector restoration was stopped and an investigation into the cause of the event was commenced. The issue! was entered into the CAP as CR 2010-0192.

Analysis.

The inspectors detenrnined that NMPNS's failure to identify the pressure detector configuration associated with 2RHS*PDT24C, and resultant failure to develop an adequate work procedure for filling and venting the detector, was a performance deficiency. The finding was more than minor because it was aSSOCiated with the procedure quality attribute of the Initiating Events cornerstone and affected the comerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.

Additionally, the finding was similar to example 4.b in Appendix E of IMC 0612, in that it resulted in a reactor scram. The inspectors evaluated the significance of this finding using IMC 0609, Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings." The finding was of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The finding had a cross-cutting aspect in the area of human perfonrnance, resources, because NMPNS did not provide maintenance personnel with an accurate work package for filling and venting the 'C' RHR pressure detector (H.2.c per IMC 0310).

Enforcement.

TS 5.4, "Procedures," states, in part, "Written procedures shan be established, implemented, and maintained covering ... the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 ..."

Regulatory Guide 1,33, Revision 2, Appendix A, February 1978, Section 9, "Procedures for Performing Maintenance," states, in part, "Maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures ... appropriate to the circumstances."

Contrary to the above, on January 7, 2010, the written procedure that was used to fill and vent RHR differential pressure detector 2RHS*PDT24C was not appropriate to the circumstance because it did not account for the potential response of other in-service pressure detectors, connected to 2RHS"PDT24C through a common reference line, to this operation. As a result, during the RHR venting evolution, an invalid transient low reactor vessel water level condition was sensed by pressure detector 2ISC*PT4B, which actuated the ARI function of the RRCS and produced a reactor scram. Because this violation was of very low safety significance and was entered into the CAP as CR 2010 0192, this violation is being treated as an NCV, consistent with the NRC Enforcement Policy. (NeV 05000410/2010002-03, Reactor Scram due to Inadequate Procedure for RHR Detector Restoration)40A5 Other Activities

.1 (Closed) Temporary Instruction 2515/177* Managing Gas Accumulation in Emergency

Core Cooling. Decay Heat Removal and Containment Spray Systems

a. Inspection Scope

The inspectors performed this inspection in accordance with Temporary Instruction (TI) 2515/177, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal and Containment Spray Systems," for NMPNS, Units 1 and 2. The NRC staff developed TI 2515/177 to support the NRC's confirmatory review of licensees' response to NRC GL 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal and Containment Spray Systems." The Office of Nuclear Reactor Regulation (NRR) documented completion of their review of NMPNS's GL 2008-01 response in a closure letter dated December 9, 2009 (ADAMS Accession No.

ML093410026). Based on the review of NMPNS's GL 2Q08-01 response letters, the NRR staff provided guidance on TI inspection scope to the regional inspectors. The inspectors used this inspection guidance along with the TI to verify that NMPNS implemented, or was in the process of acceptably implementing, the commitments, modifications, and programmatically controlled actions described in their GL 2008-01 response. The inspectors verified that the plant-speCific information (including licensing bases documents and design information) was consistent with the information used by NRR in their assessment and that it supported a conclusion that the subject systems' operability was reasonably assured.

The inspectors reviewed a sample of isometric drawings and piping and instrument diagrams, and conducted selected system piping walkdowns to verify that NMPNS had drawings that reflected the subject system configurations and UFSAR deSCriptions.

Specifically, the inspectors verified the following related to a sample of isometric drawings for the Unit 1 core spray and containment spray systems, and for the Unit 2 LPCS, HPCS and RHR systems:

• High point vents were identified;

• High points that did not have vents were recognized and evaluated with respect to their potential for gas buildup;

• Other areas where gas could accumulate and potentially impact subject system operability, such as at orifices in horizontal pipes, isolated branch lines, heat exchangers, improperly sloped piping, and under closed valves, were acceptably evaluated in engineering reviews or had ultrasonic testing (UT) points which would reasonably detect void formation; and

• For piping segments reviewed, branch lines and fittings were clearly shown.

The linspectors conducted walkdowns of portions of the above systems to reasonably assure the acceptability of NMPNS's drawings utilized during their review of the GL.

The inspectors verified that NMPNS conducted walkdowns of the above systems to confirm that system orientations, vents and alarms, in combination with instructions, procedures, tests, and training, would ensure that each system was sufficiently full of water to assure operability. The inspectors reviewed NMPNS's methodology for the determination of system piping high points, identification of negative sloped piping, and calculations of void sizes based on UT equipment readings, to ensure NMPNS's methods were reasonable. The inspectors reviewed engineering analyses associated with the development of acceptance criteria for as-found voids within the Unit 2 LPCS and 'C' RHR system suction piping. The review included engineering assumptions for void transport and acceptability of void fractions at the inlet of the pumps. The inSpEtctors also performed a walkdown with NMPNS personnel to observe where field UT measurement locations had been utilized for the monitoring of gas voids within the Unit :2 'C' RHR and LPCS pump suction piping to assess the adequacy of the monitoring plan used to ensure system operability.

The inspectors reviewed a sample of NMPNS's procedures used for filling and venting the associated GL 2008-01 systems to verify that the procedures were effective in venting or reducing voiding to acceptable levels. The inspectors verified that NMPNS's specified surveillance frequencies were consistent with NMPNS's TSs, TS bases, and the UFSAR. The inspectors reviewed a sample of system STs, regarding system venting, to ensure procedures were revised to document the existence of as-found gas conditions and evaluate within the CAP as necessary. The inspectors reviewed CAP documents to verify that selected actions described in NMPNS's nine-month and supplemental submittals were acceptably documented including completed actions and the implementation schedule for incomplete actions. Additionally, the inspectors reviewed NMPNS's evaluations and corrective actions for various issues identified durin,~ their GL 2008-01 review. This reView was performed to ensure NMPNS appropriately evaluated and adequately addressed any gas voiding concerns, including the evaluation of operability for gas voids discovered in the field. The inspectors interviewed NMPNS training personnel to assess if appropriate training had been provided to the operations staff to ensure appropriate awareness of the effects of gas voiding. The inspectors also discussed gas voiding concerns with design and system engineers, to assess their awareness of gas voiding issues and the effectiveness of NMPNS's training.

b. Findings

No findings of significance were identified. This completes the inspection requirements for thi:s TJ.

.2 (Closed) TI 2515/180 - Inspection of Procedures and Processes for Managing Fatigue

a. Inspection Scope

The objective of this TI was to determine if Constellation Energy Nuclear Group's (CENG's) implementation procedures and processes required by 10 CFR Part 26, Subpart I, "Managing Fatigue," are in place to reasonably ensure that the requirements specified in Subpart I are being addressed. This TI applies to all operating nuclear powElr reactor licensees, but is intended to be performed for one site per utility. On March 22-23, 2010, the inspector interfaced with the appropriate station staff to obtain and review station policies, procedures, and processes necessary to complete all portions of this TI.

b. Findings and Observations

No findings of Significance were identified. The inspectors confirmed that the CENG procedures listed in Section 40A5 of the Attachment contained the necessary processes to ensure compliance with requirements in 10 CFR Part 26, Subpart I, "Managing Fatigue." This completes the inspection requirements for this TI.

40A6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to Mr. Sam Belcher and other members of IiCEmsee management at the conclusion of the inspection on April 16, 2010. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

ATIACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

S. Belcher, Vice President

T. Lynch, Plant General Manager

W. Byrne, Manager, NUclear Safety and Security

J. Dean, Director Nuclear Oversight

R Dean, Training Manager

S. Dhar, Design Engineering

J. Holton, Supervisor, Systems Engineering

J. Kaminski, Director, Emergency Preparedness

J. Krakuszeski, Manager, Operations

M. Kunzwiler, Security Supervisor and Fatigue Rule Program Coordinator

F. Payne, Unit 1 General Supervisor Operations

M. Shanbhag, Licensing Engineer

S. Sova, Radiation Protection Manager

H. Strahley, Unit 2 General Supervisor Operations

T. Syrell, Director, Licensing

J. Vaughn, Operations Engineer

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None.

Opened and Closed

05000410/2010002-01 FIN Inadequate Maintenance Procedure Results in Loss of Loads for Non Vital UPS (Section 1R13)

05000410/2010002-02 NCV Inadequate Performance Testing of the Division 1 Battery (Section 1R18)

05000410/2010002-03 NCV Reactor Scram due to Inadequate Procedure for RHR Detector Restoration (Section 40A3)

Clos~g None.

Discussed

None.