IR 05000483/2009002

IR 05000483-09-002, on 01/01/2009 - 03/24/2009, Callaway Plant Integrated Resident and Regional Report, Event Follow-up

ML091250117
Person / Time
Site:	Callaway
Issue date:	04/28/2009
From:	Vincent Gaddy NRC/RGN-IV/DRP
To:	Heflin A Union Electric Co
References
IR-09-002
Download: ML091250117 (37)

Text

UNITED STATES NUC LE AR RE G UL AT O RY C O M M I S S I O N R E GI ON I V 612 EAST LAMAR BLVD , SU I TE 400 AR LI N GTON , TEXAS 76011-4125 April 28, 2009 Mr. Adam C. Heflin, Senior Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, MO 65251 Subject: CALLAWAY PLANT - NRC INTEGRATED INSPECTION REPORT 05000483/2009002

Dear Mr. Heflin:

On March 24, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Callaway Plant. The enclosed integrated inspection report documents the inspection findings, which were discussed on March 23, 2009, with Mr. T. Hermann, Vice President, Nuclear Engineering, and other 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 two findings of very low safety significance (Green). Both of these 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, the NRC is treating these findings as noncited violations, consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest the violations or the significance of the noncited violations, 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, D.C. 20555-0001, with copies to the Regional Administrator, U.S.

Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Callaway Plant facility.

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 X, and the NRC Resident Inspector at

[Plant Name]. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its enclosure, will be available electronically for public inspection in the NRC Public Document

Union Electric Company -2-Room or from the Publicly Available Records component of 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/

Vincent G. Gaddy, Chief Project Branch B Division of Reactor Projects Docket: 50-483 License: NPF-30

Enclosure:

NRC Inspection Report 05000483/2009002 w/Attachment: Supplemental Information

REGION IV==

Docket: 05000483 License: NPF-30 Report: 05000483/2009002 Licensee: Union Electric Company Facility: Callaway Plant Location: Junction Highway CC and Highway O Fulton, MO Dates: January 1 through March 24, 2009 Inspectors: D. Dumbacher, Senior Resident Inspector J. Groom, Resident Inspector P. Elkmann, Senior Emergency Preparedness Inspector Approved By: V. Gaddy, Chief, Project Branch B Division of Reactor Projects-1- Enclosure

SUMMARY OF FINDINGS

IR 05000483, 01/01-03/24/2009; Callaway Plant Integrated Resident and Regional Report;

Event Follow-up.

The report covered a 3-month period of inspection by resident inspectors and announced baseline inspections by regional based inspectors. Two Green noncited violations of significance were identified. The significance of most findings is indicated by their color (Green,

White, Yellow, or Red) using Inspection Manual Chapter 0609, Significance Determination Process. Findings for which the significance determination process 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: Initiating Events

Green.

The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after operator response to an electrical fault on the condensate Pump C motor resulted in an unplanned and unnecessary reactor trip, feedwater isolation, and auxiliary feedwater actuation. On December 11, 2008, Callaway Plant experienced an automatic turbine trip/reactor trip during a power reduction initiated by the operators response to a loss of condensate Pump C. The control room supervisor directed a power reduction without immediately referencing Procedure OTO-AE-00001 guidance and without specifying any magnitude or rate limitations on the power reduction.

The balance of plant reactor operator, not aware of the procedural limitations, initiated the power reduction using the turbine controls load limiter potentiometer. This method of turbine load control eliminated all automatic rate-limiting functions. The steam generator levels increased rapidly with sluggish main feedwater regulating valves slowing anticipatory response. The steam generator P-14 high-high level turbine trip/reactor trip occurred about 5 minutes after condensate Pump C had tripped.

This finding was greater than minor because it was associated with the Initiating Events cornerstone attribute of procedural quality and it affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. The finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively establish clear expectations and standards regarding procedurally directed actions versus actions viewed as necessary to stabilize a plant transient H.4(b)

(Section 4OA3).

Green.

The inspectors identified a self-revealing noncited violation of Technical Specification 5.4.1.a, Procedures, after maintenance on intermediate range nuclear Instrument N36 resulted in an unanticipated reactor trip signal and feedwater isolation. On December 12, 2008, Callaway instrumentation and controls maintenance personnel performed work to replace a circuit card associated with the intermediate range nuclear Instrument P-6 bistable. At the time of the maintenance, the plant was in Mode 3 with the reactor trip breakers open. Shortly after beginning work, an intermediate range high flux reactor trip signal was generated. The trip signal was generated because the bypass of the reactor trip bistables is removed upon removal of the control power fuses. With instrument power removed, the solid state protection system perceived a high intermediate range neutron flux condition and generated a reactor trip signal and feedwater isolation. Control room operators responded to the feedwater isolation by starting both motor-driven auxiliary feedwater pumps and restoring steam generator water levels to the program band. The licensee later determined that instrumentation and controls maintenance personnel were unaware that pulling the control power fuses would cause a reactor trip signal and that the step in the work instruction that directed the removal of the control power fuses had not received an adequate review.

This finding was greater than minor because the finding impacted the Initiating Events cornerstone attribute of human performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609.04, Phase 1 -

Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. This issue was entered into the licensee's corrective action program as Callaway Action Request 200812681. The finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to coordinate the impact of changes to the work scope or activity, specifically, the licensee failed to fully evaluate the impact of removal of control power fuses on the work instructions

H.3(b) (Section 4OA3).

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status

AmerenUE operated the Callaway Plant near 100 percent until February 19, 2009, when the plant was shut down, as required by the plants Technical Specifications, due to a power supply failure in the balance of plant engineered safety feature actuation system. The plant shutdown was followed by a hydrogen leak from the main generator that necessitated licensee declaration of a Notice of Unusual Event. The reactor was restarted on February 20, 2009, but the plant did not synchronize with the grid due to an additional hydrogen leak identified from the main generator. The reactor was shut down on February 21, 2009, to facilitate additional repair efforts to the main generator. Following repairs, the plant was restarted on March 1, 2009, and returned to near 100 percent power on March 4, 2009. The plant was maintained at full power for the remainder of the inspection period.

REACTOR SAFETY

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

1R01 Adverse Weather Protection

Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

On January 16, 2009, the inspectors performed a review of the licensees adverse weather procedures for seasonal extreme cold temperatures. The inspectors verified that weather-related equipment deficiencies identified during the previous year were corrected prior to the onset of seasonal extremes; and evaluated the implementation of the adverse weather preparation procedures and compensatory measures for the affected conditions before the onset of, and during, the adverse weather conditions.

During the inspection, the inspectors focused on plant-specific design features and the licensees procedures used to mitigate or respond to adverse weather conditions.

Additionally, the inspectors reviewed the Final Safety Analysis Report and performance requirements for systems selected for inspection, and verified that operator actions were appropriate as specified by plant-specific procedures. Specific documents reviewed during this inspection are listed in the attachment. The inspectors also reviewed corrective action program items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action procedures. The inspectors reviews focused specifically on the following plant systems:

• Condensate Storage Tank Freeze Protection

• Auxiliary and Fuel Building Ventilation

• Refueling Water Storage Tank Heat Trace Circuitry These activities constitute completion of one readiness for seasonal adverse weather sample as defined in Inspection Procedure 71111.01-05.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignments

.1 Partial Walkdown

a. Inspection Scope

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

• January 13, 2009, emergency diesel generator Train A while the motor-driven auxiliary feedwater pump Train B was out of service

• January 31, 2009, motor-driven auxiliary feedwater system Train A following restoration from maintenance The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Final Safety Analysis Report, Technical Specification requirements, administrative Technical Specifications, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of two partial system walkdown samples as defined in Inspection Procedure 71111.04-05.

b. Findings

No findings of significance were identified.

.2 Complete Walkdown

a. Inspection Scope

On February 11, 2009, the inspectors performed a complete system alignment inspection of essential service water system to verify the functional capability of the system. The inspectors selected this system because it was considered either safety-significant or risk-significant in the licensees probabilistic risk assessment. The inspectors walked down the system to review mechanical and electrical equipment line ups, electrical power availability, system pressure and temperature indications, as appropriate, component labeling, component lubrication, component and equipment cooling, hangers and supports, operability of support systems, and to ensure that ancillary equipment or debris did not interfere with equipment operation. The inspectors reviewed a sample of past and outstanding work orders to determine whether any

deficiencies significantly affected the system function. In addition, the inspectors reviewed the corrective action program database to ensure that system equipment-alignment problems were being identified and appropriately resolved. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one complete system walkdown sample as defined in Inspection Procedure 71111.04-05.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Quarterly Fire Inspection Tours

a. Inspection Scope

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

• January 12, 2009, fire Area A-14, Room 1326 motor-driven auxiliary feedwater pump Train A

• January 26, 2009, fire Area A-25, Room 1322, south mechanical piping penetration room

• February 5, 2009, fire Area A-8, auxiliary building 2000 elevation general area

• February 17, 2009, fire Area F-5, Room 6202, fuel building electrical equipment room 2026 elevation

• February 24, 2009, turbine building 2065 elevation

• March 2, 2009, reactor building The inspectors reviewed areas to assess if licensee personnel had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant; effectively maintained fire detection and suppression capability; maintained passive fire protection features in good material condition; and had implemented adequate compensatory measures for out of service, degraded or inoperable fire protection equipment, systems, or features, in accordance with the licensees fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to affect equipment that could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event. Using the documents listed in the attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed, that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees corrective action program.

These activities constitute completion of six quarterly fire-protection inspection samples as defined in Inspection Procedure 71111.05-05.

b. Findings

No findings of significance were identified.

.2 Annual Fire Protection Drill Observation

a. Inspection Scope

On January 26, 2009, the inspectors observed a fire brigade activation for a simulated fire caused by a fuel oil spill in the auxiliary boiler room. The observation evaluated the readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee staff identified deficiencies, openly discussed them in a self-critical manner at the drill debrief, and took appropriate corrective actions. Specific attributes evaluated were:

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

(2) proper use and layout of fire hoses;

(3) employment of appropriate fire fighting techniques;

(4) sufficient firefighting equipment brought to the scene;

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

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

(7) smoke removal operations;

(8) utilization of preplanned strategies;

(9) adherence to the preplanned drill scenario; and

(10) drill objectives.

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

These activities constitute completion of one annual fire-protection inspection sample as defined in Inspection Procedure 71111.05-05.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

a. Inspection Scope

On March 17, 2009, the inspectors observed a crew of licensed operators in the plants simulator respond to a 15 gallon per minute reactor coolant system leak followed by a steam line rupture to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• Licensed operator performance

• Crews clarity and formality of communications

• Crews ability to take timely actions in the conservative direction

• Crews prioritization, interpretation, and verification of annunciator alarms

• Crews correct use and implementation of abnormal and emergency procedures

• Control board manipulations

• Oversight and direction from supervisors

• Crews ability to identify and implement appropriate Technical Specification actions and emergency plan actions and notifications The inspectors compared the crews performance in these areas to pre-established operator action expectations and successful critical task completion requirements.

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

These activities constitute completion of one quarterly licensed-operator requalification program sample as defined in Inspection Procedure 71111.11.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

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

• January 19, 2009, trips of main condensate Pumps C and B (CAR 200812732)

• January 20, 2009, main feedwater system (CAR 200811781)

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

• Implementing appropriate work practices

• Identifying and addressing common cause failures

• Scoping of systems in accordance with 10 CFR 50.65(b)

• Characterizing system reliability issues for performance

• Charging unavailability for performance

• Trending key parameters for condition monitoring

• Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or (a)(2)

• Verifying appropriate performance criteria for structures, systems, and components classified as having an adequate demonstration of performance through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as requiring the establishment of appropriate and adequate goals and corrective actions for systems classified as not having adequate performance, as described in 10 CFR 50.65(a)(1)

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of two quarterly maintenance effectiveness samples as defined in Inspection Procedure 71111.12-05.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

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

• January 12, 2009, elevated risk during turbine-driven auxiliary feedwater pump surveillance

• February 10, 2009, elevated risk during residual heat removal Train B drain and refill

• February 19, 2009, elevated risk due to balance of plant engineered safety feature actuation system Cabinet SA036D power supply failure

• March 18, 2009, elevated risk due to component cooling water system Train B maintenance The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4)and that the assessments were accurate and complete. When licensee personnel performed emergent work, the inspectors verified that the licensee personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed the Technical Specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four maintenance risk assessments and emergent work control inspection samples as defined in Inspection Procedure 71111.13-05.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• January 22, 2009, emergency diesel generator Train B jacket water heat exchanger tube thinning (CAR 200900231)

• January 27, 2009, flaw discovered in containment liner (CARs 200811479 and 200900316)

• January 29, 2009, jacket water gasket leak on emergency diesel generator Train B (CAR 200812985)

• January 29, 2009, auxiliary feedwater Check Valve ALV0009 backleakage (CAR 200900670)

• February 5, 2009, five drop per minute leak from essential service water Train A high density polyethylene flange (CAR 200900908)

• March 2, 2009, balance of plant engineered safety feature actuation system Cabinet SA036D power supply excessive ripple (CAR 200901743)

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

These activities constitute completion of six operability evaluations inspection samples as defined in Inspection Procedure 71111.15-05

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

a. Inspection Scope

The inspectors reviewed the following temporary/permanent modifications to verify that the safety functions of important safety systems were not degraded:

• January 27, 2009, permanent modification, containment spray fill and vent, MP 08-0016

• February 18, 2009, temporary modification, temporary air compressor and air dryer in the KA system, TM 09-0002

• March 2, 2009, permanent modification, approve clamp (EJ04R014) to remain in place on EJ-008-BCA-12, MP 08-0050 The inspectors reviewed key affected parameters associated with energy needs, materials/replacement components, timing, heat removal, control signals, equipment protection from hazards, operations, flow paths, pressure boundary, ventilation boundary, structural, process medium properties, licensing basis, and failure modes.

The inspectors verified that modification preparation, staging, and implementation did not impair emergency/abnormal operating procedure actions, key safety functions, or operator response to loss of key safety functions; postmodification testing will maintain the plant in a safe configuration during testing by verifying that unintended system interactions will not occur, systems, structures and components performance characteristics still meet the design basis, the appropriateness of modification design assumptions, and the modification test acceptance criteria will be met; and licensee personnel identified and implemented appropriate corrective actions associated with permanent plant modifications. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of two samples for permanent plant modifications and one temporary plant modification sample as defined in Inspection Procedure 71111.18-05

b. Findings

No findings of significance were identified.

1R19 Postmaintenance Testing

a. Inspection Scope

The inspectors reviewed the following postmaintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• January 8, 2009, refueling water storage tank Level Transmitter 930 replacement (Job 09000137)

• January 13, 2009, motor-driven auxiliary feedwater pump Train B (Job 07511953)

• February 10, 2009, control room air conditioning unit Train B and pressurization fan (Job 07511953)

• February 20, 2009, balance of plant engineered safety feature actuation system Cabinet SA036D power supply replacement (Job 09001256)

• February 21, 2009, main generator hydrogen seal (Job 09001270)

• February 24, 2009, Battery Charger NK21 postmaintenance test (Job 09001421)

• March 11, 2009, Charging Pump B postmaintenance test (Job 08514382)

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

• The effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed

• Acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate The inspectors evaluated the activities against the Technical Specifications, the Final Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with postmaintenance tests to determine whether the licensee was identifying problems and entering them in the corrective action program and that the problems were being corrected commensurate with their importance to safety. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of seven postmaintenance testing inspection samples as defined in Inspection Procedure 71111.19-05.

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

The inspectors evaluated outage activities for an unscheduled outage that began on February 19, 2009, because of a power supply failure in balance of plant engineered safety feature actuation system Cabinet SA036D and continued through March 3, 2009, due to required repairs to address a hydrogen leak from the main generator. The inspectors reviewed activities to ensure that the licensee considered risk in developing, planning, and implementing the outage schedule. As part of the their review, the inspectors observed or reviewed the reactor shutdown, outage equipment configuration and risk management, electrical lineups, selected clearances, control and monitoring of decay heat removal, control of containment activities, startup activities, and identification and resolution of problems associated with the outage. Additionally, the inspectors reviewed the extent of condition and apparent cause for the power supply failure on balance of plant engineered safety feature actuation system Cabinet SA036D and the hydrogen leak from the main generator. Specific documents reviewed during this inspection are listed in the attachment.

This inspection constitutes one other outage sample as defined in Inspection Procedure 71111.20-05

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the Final Safety Analysis Report, procedure requirements, and Technical Specifications to ensure that the six surveillance activities listed below demonstrated that the systems, structures, and/or components tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the significant surveillance test attributes were adequate to address the following:

• Preconditioning

• Evaluation of testing impact on the plant

• Acceptance criteria

• Test equipment

• Procedures

• Jumper/lifted lead controls

• Test data

• Testing frequency and method demonstrated Technical Specification operability

• Test equipment removal

• Restoration of plant systems

• Fulfillment of ASME Code requirements

• Updating of performance indicator data

• Engineering evaluations, root causes, and bases for returning tested systems, structures, and components not meeting the test acceptance criteria were correct

• Reference setting data

• Annunciators and alarms setpoints The inspectors also verified that licensee personnel identified and implemented any needed corrective actions associated with the surveillance testing.

• January 12, 2009, turbine-driven auxiliary feedwater pump and valve strokes, Jobs 08509875, 07510704, 08509875, 08511504, and 08514271

• January 21, 2009, Procedure OSP-NE-0001B, emergency diesel generator Train B surveillance, Job 09500511

• January 26, 2009, containment spray Train A inservice test, Job 08512784

• February 13, 2009, residual heat removal Train B pump inservice test, Job 08514516

• February 27, 2009, Procedure OSP-SF-00001, shutdown margin calculation

• March 18, 2009, Procedure OSP-BB-00009, reactor coolant system inventory balance (leakrate)

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

These activities constitute completion of six surveillance testing inspection samples (three routine, two inservice test, and one reactor coolant system leakage) as defined in Inspection Procedure 71111.22-05.

b. Findings

No findings of significance were identified.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The inspectors performed an in-office review of Revision 33 to the Callaway Plant Radiological Emergency Response Plan, and Revision 45 to Emergency Plan Implementing Procedure ZZ-00101, AClassification of Emergencies,@ both submitted December 8, 2008. These revisions implemented an emergency action level scheme based on Nuclear Energy Institute Report 99-01, AMethodology for Development of Emergency Action Levels,@ Revision 5, as approved in a Safety Evaluation Report dated October 3, 2008 (ADAMS Accession Numbers ML081580257 and ML0822003670).

These revisions were compared to their previous revision, to the criteria of NUREG-0654, ACriteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,@ Revision 1, to the criteria of Nuclear Energy Institute Report 99-01, AMethodology for Development of Emergency Action Levels,@ Revision 5, and to the standards in 10 CFR 50.47(b) to determine if the revisions adequately implemented the requirements of 10 CFR 50.54(q).

This review was not documented in a Safety Evaluation Report and did not constitute an approval of the licensees changes; therefore, these revisions are subject to future inspection.

These activities constitute completion of two samples as defined in IP 71114.04-05.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

.1 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the data submitted by the licensee for the fourth quarter 2008 performance indicators for any obvious inconsistencies prior to its public release in accordance with Inspection Manual Chapter 0608, Performance Indicator Program.

This review was performed as part of the inspectors normal plant status activities and, as such, did not constitute a separate inspection sample.

b. Findings

No findings of significance were identified.

.2 Mitigating Systems Performance Index - Emergency AC Power System

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance Index - Emergency AC Power System performance indicator for the period from the first quarter 2008 through the fourth quarter 2008. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees operator narrative logs, Callaway action requests, mitigating systems performance index derivation reports, event reports, and NRC integrated inspection reports for the period of first quarter 2008 through the fourth quarter 2008 to validate the accuracy of the submittals. The inspectors reviewed the mitigating systems performance index component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified.

Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one mitigating systems performance index emergency AC power system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.3 Mitigating Systems Performance Index - Cooling Water Systems

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance Index - Cooling Water Systems performance indicator first quarter 2008 through the

fourth quarter 2008. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees operator narrative logs, Callaway action requests, mitigating systems performance index derivation reports, event reports, and NRC integrated inspection reports for the period of first quarter 2008 through the fourth quarter 2008 to validate the accuracy of the submittals. The inspectors reviewed the mitigating systems performance index component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one mitigating systems performance index cooling water system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.4 Reactor Coolant System Specific Activity

a. Inspection Scope

The inspectors sampled licensee submittals for the Reactor Coolant System Specific Activity performance indicator first quarter 2008 through the fourth quarter 2008. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees reactor coolant system chemistry samples, Technical Specification requirements, Callaway action requests and event reports to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. In addition to record reviews, the inspectors observed a chemistry technician obtain and analyze a reactor coolant system sample. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one reactor coolant system specific activity sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

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

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees corrective action program at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. The inspectors reviewed attributes that included: the complete and accurate identification of the problem; the timely correction, commensurate with the safety significance; the evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrences reviews; and the classification, prioritization, focus, and timeliness of corrective actions. Minor issues entered into the licensees corrective action program because of the inspectors observations are included in the attached list of documents reviewed.

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

b. Findings

No findings of significance were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees corrective action program. The inspectors accomplished this thorough review of the stations daily corrective action documents.

The inspectors performed these daily reviews as part of their daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings of significance were identified.

.3 Selected Issue Follow-up Inspection

a. Inspection Scope

During a review of items entered in the licensees corrective action program, the inspectors focused on corrective actions associated with:

• The licensees actions to address aging management of balance of plant engineered safety feature actuation system power supplies. The inspectors identified several corrective action items, CARs 200901391, 200901689 and 200901694, documenting degraded power supplies for the balance of plant engineered safety feature actuation system that may be an aging management concern.

• The licensees actions to address a degraded main generator hydrogen seal.

The inspectors identified several corrective action items, CARs 200700791, 200701950, 200911845, 200801680 and 200812057 documenting increased main hydrogen usage. Following the main generator hydrogen leak and declaration of a Notice of Unusual Event on February 19, 2009, the inspectors reviewed the licensees cause determination, repair activities and follow-up corrective documented in CARs 200901407, 200901589, 200901639 and 200901642. The inspectors verified that the licensees corrective actions were adequate to ensure the reliability of the main generator hydrogen seals and reduce the likelihood of a follow-up plant transient.

These activities constitute completion of two in-depth problem identification and resolution samples as defined in Inspection Procedure 71152-05.

b. Findings

No findings of significance were identified.

4OA3 Event Follow-up

.1 (Closed) Licensee Event Report (LER) 05000483/2008006-00, Reactor Trip During

Power Reduction Following Loss of Condensate Pump C

a. Inspection Scope

On December 11, 2008, Callaway Plant experienced an automatic turbine trip/reactor trip during a power reduction initiated by the operators response to a loss of condensate Pump C. Procedure OTO-AE-00001, Feedwater System Malfunction, addressed loss of condensate pumps. The procedure required a power reduction of about 10% to maintain main feedwater pump suction pressures. The control room supervisor directed the power reduction without specifying any magnitude or rate limitations. The balance of plant reactor operator, not aware of the procedural limitations, initiated the power reduction using the turbine controls load limiter potentiometer. The turbine load was reduced 33 percent from an initial 1314 MWe to 840 MWe in four minutes and resulted in control rods inserting at a maximum speed and the opening of all four groups of steam dumps. The power reduction caused steam generator water levels to decrease significantly. With the steam generator levels low, the main feedwater regulating valves opened fully and the feedwater pumps automatically increased speed. This action resulted in the steam generator levels increasing rapidly with slow responding main

feedwater regulating valves. The steam generator P-14 high-high level turbine trip/reactor trip and automatic actuation of the auxiliary feedwater system occurred about 5 minutes after condensate Pump C had tripped. The inspectors reviewed the licensees root cause analysis and proposed corrective actions. The inspectors also reviewed this LER and found that the licensee adequately documented the summary of the event including the potential safety consequences, causes of the event and corrective actions required to address the performance deficiency. The performance deficiency associated with the response to the plant transient initiated by the loss of condensate Pump C is documented below. This LER is closed.

b. Findings

Introduction.

A self-revealing Green noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after operator response to an electrical fault on condensate Pump C motor resulted in an unplanned and unnecessary reactor trip, feedwater isolation, and auxiliary feedwater actuation.

Description.

On December 11, 2008, Callaway Plant experienced an automatic turbine trip/reactor trip during a power reduction initiated by the operators response to a loss of condensate Pump C. Procedure OTO-AE-00001, Feedwater System Malfunction, addressed the loss of condensate pumps. The procedure required a power reduction, limited to less than 5 percent power decrease per minute to restore main feedwater pump suction pressures if suction pressure was lost. Original diagnosis of the condensate pump trip by the operating crew was delayed because the crew assumed there was a main turbine control problem. The main generator megawatt output had increased due to less extraction steam being withdrawn from the main turbine.

Once the condensate pump trip was recognized, the control room supervisor directed a power reduction without immediately referencing Procedure OTO-AE-00001 guidance and without specifying any magnitude or rate limitations on the power reduction. The balance of plant reactor operator, not aware of the procedural limitations, initiated the power reduction using the turbine controls load limiter potentiometer. This method of turbine load control eliminated all automatic rate-limiting functions. The turbine load was reduced 33 percent in 4 minutes. The rate of load reduction resulted in control rods inserting at a maximum speed of 72 steps per minute and the opening of all four groups of steam dumps valves.

The balance of plant operator was concerned about the possibility of the main feedwater pumps losing suction pressure. He did not, however, understand that the turbine power reduction was actually worsening the feedwater pumps suction pressure and thus continued the power reduction from an initial 1314 MWe to 840 MWe in an effort to ensure adequate suction to the feedwater pumps. Steam generator water levels had decreased significantly due to the shrink effect from the power reduction and the reduced feedwater pump flows. The reactor operators belief that they had to remove a windup feature of the control logic for the main feedwater regulating valves led to placing the controllers in manual. The licensees root cause team determined that an incorrect gain setting for the main feedwater regulating valves controllers also affected the response time of the valves. With the steam generator levels low due to lack of feedwater, the main feedwater regulating valves opened fully and the feedwater pumps automatically increased speed. These actions resulted in the steam generator levels increasing rapidly with sluggish main feedwater regulating valves slowing anticipatory response. The steam generator P-14 high-high level turbine trip/reactor trip occurred about 5 minutes after condensate Pump C had tripped.

The licensees root cause evaluation identified several causal factors.

• The operator-directed actions to stabilize the plant were different from the procedurally-directed actions. The control room supervisor chose to allow the reactor operators to perform actions to stabilize the plant based on their knowledge versus verbatim compliance with the off-normal procedure.

• There was a lack of procedural guidance and a lack of reactor operator experience associated with use of the load limiter potentiometer for control of turbine load changes.

• Human Performance tools that could have questioned the rate of power decrease, the delay in use of the off-normal procedure, and the manual control of the main feedwater regulating valves were not used.

• The operating crew did not function as a team.

• There were fundamental misunderstandings of the operation of the turbine load controllers, the main feedwater regulating valve controllers, and the parameters affecting the feedwater pump suction pressure.

Analysis.

The performance deficiency associated with this finding involved the licensed operators failure to follow procedures designed to respond to degraded equipment and to address plant transients. This finding was greater than minor because the performance deficiency was associated with the initiating events cornerstone human performance attribute and it affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations.

Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. The finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively establish clear expectations and standards regarding procedurally directed actions versus actions viewed as necessary to stabilize a plant transient H.4(b).

Enforcement.

Technical Specification 5.4.1.a, Procedures, required that written procedures be established and implemented covering activities specified in Appendix A, Typical Procedures for Pressurized Water Reactors, of Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), February 1978. Regulatory Guide 1.33, Appendix A, Section 6.j, specified that plant procedures for combating emergencies and other significant events be performed in accordance with written procedures or documented instructions appropriate to the circumstances. Contrary to the above, on December 11, 2008, the plant operators failure to use plant off-normal Procedure OTO-AE-00001, designed to mitigate the consequences of a trip of a single condensate pump resulted in an automatic reactor trip, feedwater isolation, and auxiliary feedwater actuation. Because of the very low safety significance and AmerenUEs action to place this issue in their corrective action program as CAR 200812666, this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy: NCV 05000483/2009002-01, Inadequate Response to Feedwater Transient Results in Reactor Trip.

.2 (Closed) LER 05000483/2008007-00, Intermediate Range Hi Flux Reactor Protection

System Actuation While Shutdown

a. Inspection Scope

On December 12, 2008, the licensee performed maintenance on intermediate range nuclear Instrument N36 that resulted in an unanticipated reactor trip signal and feedwater isolation. The trip signal was generated because the work document directed the removal of the control power fuses with instrument power removed. In this condition, the solid state protection system perceived a high intermediate range neutron flux condition and generated a reactor trip signal. The licensees investigation identified that instrumentation and controls maintenance personnel were unaware that pulling the control power fuses would cause a reactor trip signal. The work document that directed removal of the control power fuse did not receive an adequate review per licensee procedure and consequently did not provide adequate cautions to personnel that a reactor trip signal could be generated. The inspectors reviewed the licensees root cause analysis and proposed corrective actions. The inspectors also reviewed this LER and found that the licensee adequately documented the summary of the event including the potential safety consequences, causes of the event, and corrective actions required to address the performance deficiency. The performance deficiency associated with the unanticipated reactor trip signal and feedwater isolation is documented below. This LER is closed.

b. Findings

Introduction.

A self-revealing Green noncited violation of Technical Specification 5.4.1.a, Procedures, was identified after maintenance on intermediate range nuclear Instrument N36 resulted in an unanticipated reactor trip signal and feedwater isolation.

Description.

On December 12, 2008, Callaway instrumentation and controls maintenance personnel performed work to replace a circuit card associated with the intermediate range nuclear Instrument P-6 bistable. At the time of the maintenance, the plant was in Mode 3 with the reactor trip breakers open. The work instructions were developed to initially place the intermediate range drawer in test in accordance with Procedure ISL-SE-00N36, Loop-Nuclear; Nuclear Instrument Intermediate Range N36.

This procedure is typically not used for card replacement and does not fully deenergize the intermediate range drawer. Since the desired practice at Callaway is to deenergize equipment prior to circuit card replacement to protect sensitive electronic circuits, the instrumentation and controls maintenance supervisor determined that it would be permissible to remove control power fuses to fully deenergize the equipment. Acting as the planner for the job, the instrument maintenance supervisor added an additional step to the work package to remove the control power fuses after removal of the instrument power fuses.

Shortly after beginning work, an intermediate range high flux reactor trip signal was generated. The trip signal was generated because the bypass of the reactor trip bistables is removed upon removal of the control power fuses. With instrument power removed, the solid state protection system perceived a high intermediate range neutron flux condition and generated a reactor trip signal. Since average coolant temperature was below 564°F, the trip signal resulted in a feedwater isolation signal and isolation of flow to all four steam generators. Control room operators responded to the feedwater isolation by starting both motor-driven auxiliary feedwater pumps and restoring steam generator water levels to the program band. The licensee later determined that

instrumentation and controls maintenance personnel were unaware that pulling the control power fuses would cause a reactor trip signal. The change implemented by the instrumentation and controls maintenance supervisor did not receive an adequate review per Procedure APA-ZZ-00322, Integrated Work Management Process Description, and consequently did not provide adequate cautions to personnel that a reactor trip signal could be generated by pulling the control power fuses.

Analysis.

The performance deficiency associated with this finding involved the licensees failure to perform an adequate review for change to a work package. This finding is similar to Example 4.b in Manual Chapter 0612, Appendix E, "Examples of Minor Issues," in that the error caused by the inadequate review of the work package resulted in a reactor trip signal and a feedwater system transient. The finding was determined to be greater than minor because it impacted the Initiating Events cornerstone attribute of human performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance since it did not affect the Technical Specification limit for reactor coolant system leakage or mitigation systems safety function, did not contribute to both the likelihood of a reactor trip and mitigation equipment or functions not being available, and did not increase the likelihood of a fire or internal/external flooding. The finding had a crosscutting aspect in the area of human performance associated with the work controls component because the licensee failed to coordinate the impact of changes to the work scope or activity, specifically, the licensee failed to fully evaluate the impact of removal of control power fuses on the work instructions H.3(b).

Enforcement.

Technical Specification 5.4.1.a, Procedures, required that written procedures be established and implemented covering activities specified in Appendix A, Typical Procedures for Pressurized Water Reactors, of Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), February 1978. Regulatory Guide 1.33, Appendix A, Section 9, specifies that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures or documented instructions appropriate to the circumstances.

Contrary to the above, on December 12, 2008, Job 08008675/605 was inadequate in that it inappropriately directed removal of the control power fuses for intermediate range nuclear Instrument N36 resulting in an unanticipated reactor trip signal and feedwater isolation. Because of the very low safety significance and AmerenUEs action to place this issue in their corrective action program as CAR 200812681, this violation is being treated as a noncited violation in accordance with Section VI.A.1 of the Enforcement Policy: NCV 05000483/2009002-02, Intermediate Range Hi Flux Reactor Protection System Actuation While Shutdown.

.3 (Closed) LER 05000483/2008008-00, Reactor Manually Tripped due to Condensate

Pump B Tripping due to a Motor Ground Fault On December 14, 2008, Callaway Plant experienced an electrical fault and trip of main condensate Pump B. Because of a previous condensate pump trip that occurred on December 12, 2008, the licensee was left with only one condensate pump running. This necessitated a manual reactor trip and plant shutdown which was completed by the plant operators. The licensee determined that the cause of the condensate Pump B trip was a turn to turn winding fault that progressed to a ground fault. The inspectors responded to the plant and discussed the reactor trip with operations, engineering, and licensee management personnel to gain an understanding of the event and assess follow-up

actions. The inspectors reviewed operator actions taken in accordance with licensee procedures and reviewed unit and system indications to verify that actions and system responses were as expected. The inspectors discussed the reactor trip with the licensees root cause analysis team and assessed the teams actions to gather, review, and assess information leading up to and following the reactor trip. The inspectors reviewed the initial investigation report to assess the detail of review and adequacy of the root cause and proposed corrective actions prior to unit restart. The inspectors also reviewed the initial licensee notification to verify that it met the requirements specified in NUREG-1022, "Event Reporting Guidelines." The inspectors reviewed this LER and no findings of significance were identified. This LER is closed.

.4 Notice of Unusual Event due to Hydrogen Leak from the Main Generator

a. Inspection Scope

On February 19, 2009, the licensee declared an Unusual Event because of a hydrogen leak from the main generator. Because of a previous plant shut down, the generator was not in operation at the time of the leak. Operators successfully purged the main generator of hydrogen to eliminate any hazard. The Unusual Event was terminated on February 19, 2009, after confirmation that the hazard was no longer present. The inspectors responded to the control room and verified the licensees actions were appropriate to address the leak and that appropriate notifications were initiated.

Additionally, the inspectors reviewed the cause, impact, and corrective actions associated with the hydrogen leak from the main generator.

b. Findings

No findings of significance were identified.

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors performed observations of security force personnel and activities to ensure that the activities were consistent with Callaway Plant security procedures and regulatory requirements relating to nuclear plant security.

These observations took place during both normal and off-normal plant working hours.

These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors normal plant status review and inspection activities.

b. Findings

No findings of significance were identified.

4OA6 Meetings

Exit Meeting Summary

On January 5, 2009, the emergency preparedness inspector conducted a telephonic exit meeting to present the results of the in-office inspection of changes to the licensees emergency

plan and implementing procedures to Mr. K. Bruckerhoff, Supervisor, Emergency Preparedness, who acknowledged the findings.

On March 23, 2009, the inspectors presented the inspection results to Mr. T. Hermann, Vice President, Nuclear Engineering, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors understood and acknowledged that any proprietary information reviewed would not be retained following report issuance.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

G. Bradley, Manager, Operations

K. Bruckerhoff, Supervisor, Emergency Preparedness

T. Elwood, Supervising Engineer, Regulatory Affairs/Licensing

T. Hooper, Nuclear Test Engineer

L. Kanuckel, Manager, Quality Assurance

S. Maglio, Assistant Manager, Regulatory Affairs

K. Mills, Manager, Plant Engineering

B. Pae, System Engineer

S. Petzel, Engineer, Regulatory Affairs

J. Pitts, Engineer

R. Wissel, System Engineer

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

Inadequate Response to Feedwater Transient Results in

05000483/2009002-01 NCV Reactor Trip (Section 4OA3)

Intermediate Range Hi Flux Reactor Protection System

05000483/2009002-02 NCV Actuation While Shutdown (Section 4OA3)

Closed

Reactor Trip During Power Reduction Following Loss of

05000483/2008006-00 LER Condensate Pump C (Section 4OA3)

Intermediate Range Hi Flux Reactor Protection System

05000483/2008007-00 LER Actuation While Shutdown (Section 4OA3)

Reactor Manually Tripped due to Condensate Pump B

05000483/2008008-00 LER Tripping due to a Motor Ground Fault (Section 4OA3)

Attachment