IR 05000220/2008002

IR 05000220-08-002, 05000410-08-002; 01/01/08 - 03/31/08; Nine Mile Point Nuclear Station, Units 1 and 2; Surveillance Testing

ML081270471
Person / Time
Site:	Nine Mile Point
Issue date:	05/05/2008
From:	Glenn Dentel Reactor Projects Branch 1
To:	Polson K Nine Mile Point
Dentel, G RGN-I/DRP/BR1/610-337-5233
References
IR-08-002
Download: ML081270471 (31)

Text

UNITED STATES May 5, 2008

SUBJECT:

NINE MILE POINT NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000220/2008002 and 05000410/2008002

Dear Mr. Polson:

On March 31, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Nine Mile Point Nuclear Station, Units 1 and 2. The enclosed integrated inspection report documents the inspection results discussed on April 11, 2008, with you and members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions 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 self-revealing finding of very low safety significance (Green). The finding was determined to involve a violation of NRC requirements. However, because of its very low safety significance and because it was entered into your corrective action program (CAP), the NRC is treating this finding as a non-cited violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you contest the non-cited violation noted in this report, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-001; and the NRC Senior Resident Inspector at Nine Mile Point Nuclear Station.

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 NRCs document system (ADAMS). ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Glenn T. Dentel, Chief Projects Branch 1 Division of Reactor Projects Docket No.: 50-220, 50-410 License No.: DPR-63, NPF-69 Enclosure: Inspection Report 05000220/2008002 and 05000410/2008002 w/Attachment: Supplemental Information cc w/encl:

M. Wallace, President, Constellation Generation B. Barron, Senior Vice President and Chief Nuclear Officer C. Fleming, Esquire, Senior Counsel, Constellation Energy Group, LLC M. Wetterhahn, Esquire, Winston and Strawn T. Syrell, Director, Licensing, Nine Mile Point Nuclear Station P. Tonko, President and CEO, New York State Energy Research and Development Authority J. Spath, Program Director, New York State Energy Research and Development Authority P. D. Eddy, Electric Division, NYS Department of Public Service C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law Supervisor, Town of Scriba T. Judson, Central NY Citizens Awareness Network D. Katz, Citizens Awareness Network

SUMMARY OF FINDINGS

IR 05000220/2008002, 05000410/2008002; 01/01/08 - 03/31/08; Nine Mile Point Nuclear Station,

Units 1 and 2; Surveillance Testing.

The report covered a three-month period of inspection by resident inspectors and regional specialist inspectors. One Green non-cited violation (NCV) was identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). 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 NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green.

A self-revealing, non-cited violation (NCV) of Technical Specification (TS) 5.4,

"Procedures," was identified on January 14, 2008, when technicians improperly performed a surveillance procedure which resulted in isolation of the Unit 2 RCIC system. Specifically, while performing a test of the area temperature instruments that provide high temperature isolation signals for the main steam system, technicians erroneously disconnected an electrical lead associated with the RCIC leak detection system. This resulted in an automatic isolation of the RCIC system steam supply and the unavailability of RCIC for approximately four hours. Operators immediately recognized the error and halted the surveillance procedure. Technicians reconnected the lead and operators restored RCIC to a normal standby lineup.

The finding was greater than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance in accordance with IMC 0609,

Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., "Determining the Significance of Reactor Inspection Findings for At-Power Situations," based on a Phase 3 analysis. The Region I senior reactor analyst (SRA)used the Nine Mile Point Unit 2 Standardized Plant Analysis Risk (SPAR) model and the actual out-of-service time to determine the risk significance. This finding had a cross-cutting aspect in the area of human performance because of the ineffective use of human error prevention techniques (H.4.a per IMC 0305). (Section 1R22)

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status

Nine Mile Point Unit 1 was operated at full rated thermal power (RTP) throughout the inspection period, with the exception of planned power reductions and recoveries for planned reactor recirculation pump maintenance, control rod testing, and main turbine valve testing.

Nine Mile Point Unit 2 began the inspection period at full RTP. Operators performed several planned power reductions and recoveries for control rod pattern adjustments, main turbine and main steam isolation valve testing, and control rod testing. On March 22, the reactor was shut down to commence refueling outage (RFO)

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R04 Equipment Alignment

.1 Partial System Walkdown (71111.04 - Four samples)

a. Inspection Scope

The inspectors performed four partial system walkdowns to verify risk-significant systems were properly aligned for operation. The inspectors verified the operability and alignment 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.

Documents reviewed during this inspection are listed in the Attachment. The inspectors performed partial walkdowns of the following systems:

• Unit 2 'B' residual heat removal (RHR) system, while the Division 1 low pressure emergency core cooling systems ('A' RHR and low pressure core spray) were inoperable for planned maintenance (January 17, 2008);

• Unit 1 control room air treatment system while 112 and 121 control room chillers were out of service for corrective maintenance (February 1, 2008);

• Unit 1 core spray system 12 during inservice testing of core spray system 11 (February 26, 2008); and

• Unit 2 high pressure core spray (HPCS) system, due to it being a risk significant single train system (March 6, 2008).

b. Findings

No findings of significance were identified.

.2 Complete System Walkdown (71111.04S - One sample)

a. Inspection Scope

The inspectors performed a complete walkdown of the Unit 1 emergency cooling system to identify discrepancies between the existing equipment configuration and that specified in the design documents. During the walkdown, system drawings and operating procedures were used to determine the proper equipment alignment and operational status. The inspectors reviewed the open maintenance work orders (WO) that could affect the ability of the system to perform its functions. Documentation associated with temporary modifications, operator workarounds, and items tracked by plant engineering were also reviewed to assess their collective impact on system operation. In addition, the inspectors reviewed the condition report (CR) database to verify that equipment alignment problems were being identified and appropriately resolved. Documents reviewed during this inspection are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05Q - Six samples)

a. Inspection Scope

The inspectors toured six areas important to reactor safety at NMPNS to evaluate the stations 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. Documents reviewed for this inspection are listed in the Attachment. The areas inspected included:

• Unit 1 train 11 battery and battery board rooms;

• Unit 1 train 12 battery and battery board rooms;

• Unit 1 containment spray pump room (112, 122), reactor building (RB) 198 and 237 foot elevations;

• Unit 2 RB 175 foot elevation;

• Unit 2 RB 196 foot elevation; and

• Unit 2 steam tunnel;

b. Findings

No findings of significance were identified.

1R08 Inservice Inspection Activities (71111.08 - One sample)

a. Inspection Scope

The purpose of this inspection was to assess the effectiveness of the inservice inspection (ISI) program for monitoring degradation of the reactor coolant system (RCS) boundary, risk significant piping system boundaries, and the containment boundary. The inspectors assessed the ISI activities using the criteria specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI and applicable NRC regulatory requirements.

The inspectors selected a sample of nondestructive examination (NDE) activities for observation or review, and evaluation for compliance with the requirements of ASME Section XI. Also, the inspectors selected samples of activities associated with the repair/replacement of safety related pressure boundary components. The sample selection was based on the inspection procedure objectives, risk significance, and availability. Specifically, the inspectors focused on components and systems where degradation would result in a significant challenge to the integrity of pressure boundary components.

The inspectors performed an observation of one volumetric examination (ultrasonic) and portions of a surface examination (liquid penetrant). In addition, the inspectors performed a documentation review of a magnetic particle surface examination. The sample selection included the following:

• Ultrasonic test (UT), volumetric examination, weld # 2CSL-26-05-FW005, butt weld, pipe to penetration, core spray system;

• Magnetic particle test, surface examination, welds #300 and 301, integral attachments, lugs to pipe, main steam system; and

• Liquid penetrant test, surface examination, welds SW 95, 96, 97 and 98, integral attachments, lugs to pipe, RCS.

The inspectors performed an evaluation of work activities during a drywell entry and visually examined the condition of accessible portions of the containment liner and coatings for peeling, blistering, corrosion, mechanical damage, and other degradation mechanisms. The inspectors noted that two different coatings were apparent on various locations of the internal exposed metallic surfaces of the containment liner. The inspectors reviewed documentation which supported the coating qualification in accordance with ANSI N101.2 and that all coating had been applied in accordance with Regulatory Guide 1.54.

The inspectors reviewed portions of the in-process remote visual examination of the steam dryer. The inspectors reviewed three CRs initiated as a result of the dryer examination and noted the rejectable indications reported. The indications noted had not been identified during the previous examination (previous outage in 2006). These issues were placed in the corrective action program for engineering evaluation and disposition.

The inspectors selected for review a sample of repair/rework activities which required the development and implementation of an ASME Section XI repair plan. The inspectors reviewed documentation for the weld repair of two pressure boundary valves in ASME Class 3 systems. The inspectors reviewed the ASME Section XI plans, work scope, activity sequence, weld filler metal selection, weld procedure specifications and procedure qualification records, welder qualifications, specified non-destructive tests, acceptance criteria, and post work testing. The following samples were inspected:

• WO 04-08487-00 was initiated for the mechanical and weld repair of globe valve 2IAS-V181 in the instrument air system. The repair involved the disassembly and rebuilding of the valve. The disassembly of the valve required the removal of the body to bonnet weld to access the internals for mechanical rework of the valve seats.

Restoration of the body to bonnet weld was required following the completion of the repair and installation of the valve internals.

• WO 05-21585-00 was initiated to facilitate the removal, testing, rebuilding, inspection and re-installation by welding, into the piping system of relief valve 2WCS-RV21A in the reactor water cleanup system. It was necessary to eliminate existing installation welds in order to remove, rebuild, and test the valve. Acceptance testing of the completed valve repair and welding was specified in the repair plan. A visual examination was specified for the installation welds and a system pressure test specified to verify valve and system integrity.

No sample of a previously identified recordable indication accepted as-is for continued service from the previous and the current outage was available for review during the inspection.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program (71111.11Q - Two samples)

a. Inspection Scope

The inspectors evaluated two simulator scenarios licensed operator requalification training program. The inspectors assessed the clarity and effectiveness of communications, the implementation of appropriate actions in response to alarms, the performance 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. Documents reviewed for this inspection are listed in the

. The following scenarios were observed:

• On March 17, 2008, the inspectors observed a Unit 2 operations crew during Just In Time Training (JITT) in preparation for RFO 11. The crew performed an approach to criticality, discussed the performance of surveillance procedure N2-OSP-EGS-R004, Operating Cycle Diesel Generator Simulated Loss of Offsite Power with ECCS Division I and II, and discussed plant modifications that would be performed during the outage.

• On March 18, 2008, the inspectors observed a Unit 2 operations crew during JITT training in preparation for RFO 11. The crew performed a plant cooldown, including the transition to RHR shutdown cooling in service.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12Q - Two samples)

a. Inspection Scope

The inspectors reviewed performance-based problems and the performance and condition history of selected systems to assess the effectiveness of the maintenance program. The inspectors reviewed the systems to ensure that the stations 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 sites ability to identify and address common cause failures and to trend key parameters. Documents reviewed for the inspection are listed in the Attachment. The following two maintenance rule inspection samples were reviewed:

• Unit 1 fire protection systems due to long-standing equipment problems; and

• Unit 2 service water (SW) system due to extended unavailability of the E SW pump.

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 paragraph (a)(4) of 10 CFR Part 50.65. The inspectors reviewed equipment logs, work schedules, and performed plant tours to gain assurance that actual plant configuration matched the assessed configuration. Additionally, the inspectors verified that risk management actions for both planned and emergent work were consistent with those described in station procedures. Documents reviewed for the inspection are listed in the

.

The inspectors reviewed risk assessments for the activities listed below.

Unit 1

• Week of January 21, 2008, that included 112 containment spray quarterly surveillance, an emergent issue with the 112 containment spray raw water pump packing overheating, emergency diesel generator (EDG) 102 monthly surveillance, high drywell pressure instrument trip channel test, and a power reduction to 88 percent to return 11 reactor recirculation pump to service.

• Week of January 28, 2008, that included control rod drive (CRD) pump quarterly surveillance, liquid poison system quarterly surveillance, emergency service water pump quarterly surveillance, main steam isolation valve (MSIV) partial stroke testing, and emergent activities to troubleshoot spiking on average power range monitors (APRMs) 12 and 15, and flow oscillations on 11 reactor recirculation pump.

• Week of February 12, 2008, that included a two day maintenance period on 11 high pressure coolant injection (HPCI) system, cleaning of 11 turbine building closed loop cooling (TBCLC) heat exchanger, repair of a packing leak on emergency cooling (EC)valve IV-39-11R which rendered 11 EC inoperable for two days, maintenance on vital uninterruptable power supply (UPS) 162A, EDG raw water system quarterly surveillance, securing 11 reactor recirculation pump for maintenance on its associated motor generator, and 11 reactor recirculation flow loop calibration and flow converter calibrations.

Unit 2

• Week of January 7, 2008, that included a power reduction to 80 percent for control rod pattern adjustment, reactor vessel water level low surveillance, main steam line high flow surveillance, and investigation of increased drywell unidentified leakage concurrent with a reactor recirculation pump motor winding cooler leakage alarm.

• Week of January 21, 2008, that included a two day maintenance period for the HPCS system, HPCS system quarterly surveillance, on-line motor testing and lubrication of the A control rod drive pump, Division 3 EDG monthly surveillance, and emergent maintenance to stop makeup water leakage into the standby liquid control tank.

• Week of January 28, 2008, that included Division 2 EDG monthly surveillance, a power reduction to 70 percent for control rod sequence exchange, MSIV testing, and turbine valve testing, Division 2 standby gas treatment system inoperable for one day for filter medium sampling, C RHR system inoperable for one day for planned maintenance, C RHR system quarterly surveillance, and quarterly test of emergency core cooling systems (ECCS) actuation on high drywell pressure.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15 - Seven samples)

a. Inspection Scope

The inspectors evaluated the acceptability of the operability evaluations, the use and control of compensatory measures, and the compliance with TSs. The evaluations were reviewed using criteria specified in NRC Regulatory Issue Summary 2005-20, Revision to Guidance Formerly Contained in NRC Generic Letter 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 review included verification that the operability determinations were made as specified by Procedure CNG-OP-1.01-1002, Conduct of Operability Determinations / Functionality Assessments.

The technical adequacy of the determinations was reviewed and compared to the TSs, UFSAR, and associated design basis documents (DBDs). Documents reviewed for the inspection are listed in the Attachment. The following evaluations were reviewed:

• CR 2008-531 concerning turbine first stage bowl pressure switch calibrations at Unit 1;

• CR 2006-3751 concerning environmental qualification of Unit 1 RB emergency ventilation damper position indicating switches;

• CR 2008-1721 concerning leaking Unit 1 emergency condenser vacuum breaker valve 60.1-28;

• CR 2008-618 concerning makeup water leakage into the Unit 2 standby liquid control (SLC) storage tank;

• CR 2007-7404 concerning Unit 2 Division 2 EDG operability with a failed emergency fuel oil solenoid valve;

• CR 2008-1276 concerning identification of increased post-accident head losses associated with the Unit 2 ECCS suppression pool suction strainers; and

• CR 2008-2176 concerning out of specification resistance readings on the Unit 2 Division 1 EDG potential transformer fuse/contact linkage assembly.

b. Findings

No findings of significance were identified.

1R18 Plant Modifications (71111.18 - One sample)

a. Inspection Scope

The inspectors reviewed Unit 2 permanent modification N2-05-010, Eliminate Single Point Vulnerability for Main Steam Tunnel Cooling. The purpose was to reduce the likelihood of a high temperature main steam line isolation due to loss of ventilation in the main steam lead enclosure. The inspectors assessed the adequacy of the modification package, including post-modification testing, and verified that applicable design and licensing basis requirements were met and that design margins were not degraded by the change.

b. Findings

No findings of significance were identified.

1R19 Post Maintenance Testing (71111.19 - Five samples)

a. Inspection Scope

The inspectors reviewed the post maintenance tests 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 test results adequately demonstrated restoration of the affected safety functions. Documents reviewed for this inspection are listed in the Attachment.

• Unit 1, WO 07-10842-00 that performed maintenance on the tie breaker between non-vital 600V power board 17A and vital 600V power board 17B. The retest was performed in accordance with N1-OP-30, "4.16KV, 600V, and 480V House Service."

• Unit 1, WO 07-08535-00 that repacked emergency condenser steam line drain valve IV-39-11. The retest was performed in accordance with N1-ST-Q4, Reactor Coolant System Isolation Valves Operability Test.

• Unit 1, WO 08-02028-00 that performed maintenance on the reactor protection system motor generator MG-141 voltage regulator. The retest was performed in accordance with N1-OP-48, Motor Generator Sets.

• Unit 1, WO 07-06872-00 that repaired leaking emergency condenser vacuum breaker valve 60.1-28. The retest was performed by overflowing the emergency condenser shell using N1-ST-M2, Emergency Cooling System Makeup Tank Level Control Valves Exercising Test.

• Unit 2, WO 07-01190-00 that performed inspection of the A CRD pump motor. The retest for the circuit breaker rack-out was performed in accordance with N2-OP-30, Control Rod Drive.

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities (71111.20 - In Progress)

a. Inspection Scope

The inspectors observed and/or reviewed the following Unit 2 refueling outage activities to verify that operability requirements were met and that risk, industry experience, and previous site-specific problems were considered. The refueling outage and inspection sample were in progress at the end of the inspection period. Documents reviewed for this inspection are listed in the Attachment.

• The inspectors reviewed the outage schedule and procedures, and verified that TS-required safety system availability was maintained and shutdown risk was minimized. The inspectors verified that, when specified by NMPNS procedure NIP-OUT-01, Shutdown Safety, contingency plans existed for restoring key safety functions.

• The inspectors observed portions of the plant shutdown and cooldown on March 22, and verified that the TS cooldown rate limits were satisfied.

• Through plant tours, the inspectors verified that NMPNS maintained and adequately protected electrical power supplies to safety-related equipment and that TS requirements were met.

• The inspectors verified proper alignment and operation of shutdown cooling and other decay heat removal systems. The verification also included reactor cavity and fuel pool makeup paths and water sources, and administrative control of drain down paths.

• The inspectors verified that requirements for refueling operations were met through refuel bridge observations, control room panel walkdowns, and discussions with Operations Department personnel.

• After the drywell was open for general access, the inspectors performed an as-found walkdown to identify evidence of RCS leakage and assess the condition of drywell structures, piping, and supports.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22 - Eight 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. Documents reviewed for this inspection are listed in the Attachment.

The following STs were reviewed:

• N1-ST-M8, RB Emergency Ventilation System Operability Test;

• N1-ST-Q6C, Containment Spray System Loop 112 Quarterly Operability Test;

• N1-ST-Q21, Instrument Air Valves Quarterly Test;

• N1-ISP-201-022, Drywell Water Leak Detection Instrument Channel Test;

• N2-ISP-LDS-R106, Main Steam Line Tunnel and MSL Lead Enclosure Temperature Instrument Channel Calibration;

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

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

• N2-OSP-GTS-R001, Secondary Containment Integrity Test.

b. Findings

Introduction.

A self-revealing Green NCV of TS 5.4, "Procedures," was identified on January 14, 2008, when technicians improperly performed a surveillance procedure which resulted in isolation of the Unit 2 RCIC system. Specifically, technicians erroneously disconnected an electrical lead associated with the RCIC leak detection system, which resulted in an automatic isolation of the RCIC system steam supply.

Description.

On January 14, 2008, instrument and controls technicians were performing an ST on the area temperature instruments that provide high temperature isolation signals for the main steam system. The surveillance, N2-ISP-LDS-R106, "Main Steam Line Tunnel and MSL Lead Enclosure Temperature Instrument Channel Calibration," requires that the associated thermocouple leads be disconnected prior to performing the channel calibration. When the technicians attempted to perform this action (step 7.2.1 of 1), they incorrectly identified the specified terminals. The procedure directed the technicians to disconnect thermocouple leads from 2CEC*PNL707 Bay F, TH405 TB-1, terminals 10 and 11. The lead that was actually disconnected was from terminal 14. One of the technicians had initially questioned the adequacy of their terminal identification since the terminals were not individually labeled. However, they concluded that they had identified the correct terminal and proceeded. The wires that they proceeded to disconnect were thermocouple leads for a temperature instrument that provides area high temperature isolation for the RCIC system. The open circuit created by lifting the first lead resulted in an automatic isolation of the RCIC system steam supply.

Operators immediately recognized the error and halted the surveillance procedure.

Technicians reconnected the thermocouple, and operators restored RCIC to a normal standby lineup. During the four hours that the RCIC steam supply was isolated, the RCIC system was inoperable and unavailable. The TS allowed outage time for the RCIC system is 14 days.

The performance deficiency associated with this event was that technicians did not correctly perform a ST procedure, which caused the Unit 2 RCIC system to automatically isolate, rendering the system unavailable to perform its safety function.

Analysis.

The finding was greater than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to Initiating Events to prevent undesirable consequences.

The inspectors evaluated the significance of this finding using IMC 0609, Appendix A, Phase 1, and determined that a Phase 2 analysis was required because the finding represented an actual loss of the RCIC system safety function for four hours. The Region I SRA determined that a Phase 3 analysis was necessary because the site-specific Phase 2 notebook indicated that the finding could be more than of very low safety significance assuming an exposure time of three days. The SRA used the Nine Mile Point Unit 2 SPAR model and the actual four-hour exposure time to determine that the increase in core damage frequency was in the range of 1 core damage accident in 1.25E8 years of reactor operation, or high E-9 per year. The SPAR model dominant cutsets were a station blackout with failure of high pressure injection sources and the inability to restore AC power within 30 minutes. Based on this review, the SRA concluded that the finding was of very low safety significance (Green).

The finding had a cross-cutting aspect in the area of human performance because of the ineffective use of human error prevention techniques, in that, although peer checking had identified a question, that question was not adequately resolved prior to proceeding (H.4.a per IMC 0305)

Enforcement.

TS 5.4, Procedures, states that written procedures shall 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, Item 8, Procedures for Control of Measuring and Test Equipment and for STs, Procedures, and Calibrations, lists containment isolation tests as an applicable group of tests. Contrary to the above, Unit 2 Instrument Surveillance Procedure N2-ISP-LDS-R106, "Main Steam Line Tunnel and MSL Lead Temperature Instrument Channel Calibration," was not correctly implemented.

On January 14, 2008, while attempting to perform Procedure Attachment 1, step 7.2.1, to disconnect field thermocouple leads from 2CEC*PNL707 Bay F, TH405 TB-1, terminals 10 and 11, technicians incorrectly disconnected the lead from terminal 14. This action resulted in an automatic isolation of the RCIC system steam supply. Because this procedural noncompliance is of very low safety significance and was entered into the CAP as CR 2008-332, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000410/2008002-01, Failure to Correctly Perform Procedure Caused Inadvertent Isolation of the RCIC Steam Supply.

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation (71114.06 - One sample)

a. Inspection Scope

The inspectors completed one emergency drill evaluation inspection sample. The inspectors observed simulator, technical support center (TSC), and operations support center activities associated with the Unit 1 emergency planning drill on March 4, 2008. The scenario consisted of a leak from the spent fuel pool (SFP) due to an earthquake during the previous shift, a loss of off-site power (including power to the TSC) with failure of one main steam line to isolate, and a main steam line break outside secondary containment. The inspectors verified that emergency classification declarations and notifications were completed in accordance with 10 CFR 50.72, 10 CFR 50, Appendix E, and the Nine Mile Point emergency plan implementing procedures. Documents reviewed for this inspection are listed in the Attachment.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Control to Radiologically Significant Areas (71121.01 - Eight samples)

a. Inspection Scope

Based on the work activities during the Unit 2 refueling outage, the inspectors selected three jobs (drywell scaffold, drywell inservice inspection, and in-vessel visual inspection)being performed in radiation areas, airborne radioactivity areas, or high radiation areas

(<1 R/hr) for observation. The inspectors observed work that was estimated to result in the highest collective doses, involved diving activities in or around spent fuel or highly activated material, or that involved potentially changing (deteriorating) radiological conditions. The inspectors reviewed all radiological job requirements (radiation work permit requirements and work procedure requirements). The inspectors observed job performance with respect to these requirements. The inspectors determined if radiological conditions in the work area were adequately communicated to workers through briefings and postings.

During job performance observations, the inspectors verified 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), and contamination controls. For high radiation work areas with significant dose rate gradients (factor of 5 or more), the inspectors reviewed the application of dosimetry to effectively monitor exposure to personnel.

During job performance observations, the inspectors observed radiation worker performance with respect to stated radiation protection work requirements. The inspectors determined if workers were aware of the significant radiological conditions in their workplace and the radiation work permit controls/limits in place, and that their performance took into consideration the level of radiological hazards present.

During job performance observations, the inspectors observed radiation protection technician performance with respect to all radiation protection work requirements. The inspectors determined if they were aware of the radiological conditions in their work area and the radiation work permit controls/limits, and if their performance was consistent with their training and qualifications with respect to the radiological hazards and work activities.

The inspectors identified exposure significant work areas within radiation areas, high radiation areas (<1 R/hr), or airborne radioactivity areas in the plant and reviewed the associated controls and surveys of these areas to determine if controls (e.g., surveys, postings, barricades) were acceptable. The areas reviewed by the inspectors included the drywell, inside the bioshield, under vessel and on the refueling floor.

With a survey instrument, the inspectors walked down these areas or their perimeters to determine whether prescribed radiation work permits, procedure, and engineering controls were in place, whether surveys and postings were complete and accurate, and whether air samplers were properly located.

The inspectors reviewed radiation work permits used to access these and other high radiation areas and identified what work control instructions or control barriers had been specified. The inspectors used plant-specific TS high radiation area requirements as the standard for the necessary barriers. The inspectors reviewed electronic personal dosimeter alarm set points (both integrated dose and dose rate) for conformity with survey indications and plant policy. The inspectors verified that workers knew what actions are required when their electronic personal dosimeter noticeably malfunctions or alarms.

The inspectors evaluated performance against the requirements contained in 10 CFR Part 20, Unit 1 TS 6.7, and Unit 2 TS 6.12.

b. Findings

No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02 - Four samples)

a. Inspection Scope

The inspectors obtained a list of work activities ranked by actual/estimated exposure that were in progress during the refueling outage and selected three work activities of highest exposure significance (see section 2OS1 above).

The inspectors reviewed the ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements. The inspectors determined whether procedures, engineering and work controls had been established based on sound radiation protection principles to achieve occupational exposures that were ALARA. The inspectors determined whether the radiological work had been reasonably grouped into work activities based on historical precedence, industry norms, and/or special circumstances.

The inspectors compared the results achieved (dose rate reductions, person-rem used)with the intended dose established in ALARA planning for these work activities. The inspectors reviewed, where applicable, inconsistencies between intended and actual work activity doses.

Based on scheduled work activities and associated exposure estimates, the inspectors selected three work activities in radiation areas, airborne radioactivity areas, or high radiation areas for observation. The inspectors concentrated on work activities that presented the greatest radiological risk to workers. The inspectors evaluated use of ALARA controls for these work activities by evaluating use of engineering controls to achieve dose reductions.

The inspectors evaluated Constellations performance against the requirements contained in 10 CFR Part 20.1101.

b. Findings

No findings of significance were identified.

2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03 - One sample)

a. Inspection Scope

The inspectors identified the types of portable radiation detection instrumentation used for job coverage of high radiation area work, other temporary area radiation monitors currently used in the plant, and continuous air monitors associated with jobs with the potential for workers to receive 50 mrem committed effective dose equivalent.

The inspectors evaluated performance against the requirements contained in 10 CFR Part 20.1501, 10 CFR Part 20.1703 and 10 CFR Part 20.1704.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification (71151 - Four samples)

a. Inspection Scope

The inspectors sampled NMPNS submittals for the performance indicators (PIs) listed below. To verify the accuracy of the PI data reported during that period, the PI definition guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline," Revision 5, was used to verify the basis in reporting for each data element.

Cornerstone: Initiating Events

The inspectors reviewed licensee event reports (LERs) and operator logs to determine whether NMPNS accurately reported the number of unplanned scrams at Unit 1 and Unit 2 from July 2007 to December 2007.

• Unit 1 and Unit 2 unplanned scrams per 7000 critical hours; and

• Unit 2 and Unit 2 unplanned scrams with complications.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

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 Nine Mile Points 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.

The ISI inspectors reviewed a sample of examination reports and CRs initiated during ISI examinations to evaluate Constellations effectiveness in the identification and resolution of problems. The inspectors reviewed CRs 2008-2332, 2008-2345, and 2008-2363, which identified flaws and other nonconforming conditions discovered during this outage. The inspectors verified that the nonconforming conditions identified were reported, characterized, and entered into the CAP.

b. Findings

No findings of significance were identified.

4OA6 Meetings, including Exit

Exit Meeting Summary

The inspectors presented the inspection results to Mr. Keith Polson and other members of NMPNS management on April 11, 2008. NMPNS acknowledged that no proprietary information was involved.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

K. Polson, Vice President

S. Belcher, Plant Manager

R. Dean, Director, Quality and Performance Assessment

J. Laughlin, Manager, Engineering Services

J. Krakuszeski, Manager, Operations

J. Kaminski, Manager, Emergency Preparedness

T. Shortell, Manager, Training

S. Sova, Manager, Radiation Protection

T. Syrell, Director, Licensing

W. Byrne, Manager, Nuclear Security

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000410/2008002-01 NCV Failure to Correctly Perform Procedure Caused Inadvertent Isolation of RCIC Steam Supply (Section 1R22)

Closed

None.

Discussed

None.