Text

April 30, 2008

Mr. Tom E. Tynan Vice President - Vogtle Southern Nuclear Operating Company, InVogtle Electric Generating Plant 7821 River Road Waynesboro, GA 30830

SUBJECT: VOGTLE ELECTRIC GENERATING PLANT- NRC INTEGRATED INSPECTION REPORT 05000424/2008002 AND 05000425/2008002 AND MEETING

SUMMARY

Dear Mr. Tynan:

On March 31, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Vogtle Electric Generating Plant, Units 1 and 2. The enclosed integrated inspection report documents the inspection findings, which were discussed on April 16, 2008, with Mr. R.

Dedrickson and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents two NRC-identified findings of very low safety significance (Green) which 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 non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC

Enforcement Policy. If you contest these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory

Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to the

Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the resident inspector at the Vogtle

Electric Generating Plant.

SNC 2In accordance with the Code of Federal Regulations 10 CFR 2.390 of the NRC's Rules of Practice, a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly

Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely, /RA/

Scott M. Shaeffer, Chief

Reactor Projects Branch 2

Division of Reactor Projects

Docket Nos.: 50-424, 50-425

License Nos.: NPF-68 and NPF-81

Enclosure: Inspection Report 05000424/2008002 and 05000425/2008002 w/Attachment: Supplemental Information cc w/encl: (See page 3)

SNC 2In accordance with the Code of Federal Regulations 10 CFR 2.390 of the NRC's Rules of Practice, a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly

Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely, /RA/

Scott M. Shaeffer, Chief

Reactor Projects Branch 2

Division of Reactor Projects

Docket Nos.: 50-424, 50-425

License Nos.: NPF-68 and NPF-81

Enclosure: Inspection Report 05000424/2008002 and 05000425/2008002 w/Attachment: Supplemental Information cc w/encl: (See page 3)

G PUBLICLY AVAILABLE G NON-PUBLICLY AVAILABLE G SENSITIVE G NON-SENSITIVE ADAMS: G Yes ACCESSION NUMBER:_________________________ OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII/DRS RII:DRS RII/DRS SIGNATURE CWR1 SMS GJM1 by email TEC1 by email CAP3 by email LFL by email RMB1 NAME CRapp SShaeffer GMcCoy TChandler CPeabody LLake RBerryman DATE 04/30/2008 04/30/2008 04/27/2008 04/28/2008 04/23/2008 04/24/2008 04/30/2008 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICIAL RECORD COPY DOCUMENT NAME: I:\RPB2\VOGTLE\REPORTS\08-02\IIR 2008-002.DOC

SNC 3cc w/encl: N. J. Stringfellow

Manager

Licensing

Southern Nuclear Operating Company, Inc.

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Jeffrey T. Gasser

Executive Vice President

Southern Nuclear Operating Company, Inc.

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L. Mike Stinson

Vice President

Fleet Operations Support

Southern Nuclear Operating Company, Inc.

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Michael A. MacFarlane

Southern Nuclear Operating Company, Inc.

Inverness Center Parkway

P.O. Box 1295

Birmingham, AL 35201-1295

Laurence Bergen

Oglethorpe Power Corporation

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Bob Masse

Resident Manager

Vogtle Electric Generating Plant

Oglethorpe Power Corporation

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Resident Manager

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Alvin W. Vogtle Nuclear Plant 21 River Road

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SNC 4 Letter to Tom E. Tynan from Scott M. Shaeffer dated April 30, 2008

SUBJECT: VOGTLE - NRC INTEGRATED INSPECTION REPORT 05000424/2008 AND 05000425/2008

Distribution w/encl: C. Evans, RII

L. Slack, RII

OE Mail

RIDSNRRDIRS

PUBLIC

R. Jervey, NRR

U. S. NUCLEAR REGULATORY COMMISSION REGION II

Docket Nos.: 50-424, 50-425

License Nos.: NPF-68, NPF-81

Report Nos.: 05000424/2008002 and 05000425/2008002

Licensee: Southern Nuclear Operating Company, Inc. (SNC)

Facility: Vogtle Electric Generating Plant, Units 1 and 2

Location: Waynesboro, GA 30830

Dates: January 1, 2008 through March 31, 2008

Inspectors: G. McCoy, Senior Resident Inspector

T. Chandler, Resident Inspector

C. Peabody, Reactor Inspector (Sections 1R04 and 1R22)

L. Lake, Senior Reactor Inspector (Section 1R07) R. Berryman, Senior Reactor Inspector (Section 4OA5)

Accompanying

Personnel: B. Collins, Reactor Inspector

Approved by: Scott M. Shaeffer, Chief Reactor Projects Branch 2

Division of Reactor Projects

SUMMARY OF FINDINGS IR 05000424/2008-002, 05000425/2008-002; 01/01/2008 - 03/31/2008; Vogtle Electric

Generating Plant, Units 1 and 2; Other Activities. The report covered a three-month period of inspection by two resident inspectors and three reactor inspectors. Two Green findings were identified both of which were non-cited violations. The significance of most findings is indicated by its color (Green, White,

Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, ASignificance Determination Process@ (SDP). Findings for which the SDP does not apply may be Green or assigned a severity level after management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in

NUREG-1649, AReactor Oversight Process@.

A. NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for failure to consider the effects of several factors such as instrument uncertainty and actual water temperature during auxiliary feedwater (AFW) pump technical specification required surveillance testing. This finding was entered into the licensee's corrective action program as condition reports CR 2007105436, CR 2007105713, CR 2007105870, and CR 2007105895. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and operability of the AFW pumps to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e. maintenance and testing procedures. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in any AFW pumps being inoperable based upon a subsequent review of AFW surveillance testing results. (Section 4OA5.1)

Green.

The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to include the cumulative effects of the replacement of the 1A motor driven auxiliary feedwater (AFW) pump rotating element, accuracy of AFW system resistance values, safety relief valve setpoint tolerances, and turbine driven

AFW pump speed settings on evaluation of the performance of the AFW system. This finding was entered into the licensee's corrective action program as condition report

CR 2007105979. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and operability of the AFW system to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance 3(Green) because the deficiencies did not result in the AFW system being inoperable based upon subsequent analysis that showed that the AFW system had sufficient flow performance margin to accommodate pump performance and the increased system flow resistance when applying appropriate resistance values and steam generator backpressures. (Section 4OA5.2)

B. Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status

Unit 1 started the period at essentially full rated thermal power (RTP). The unit was shutdown on March 16 for a planned refueling outage. Unit 2 operated at essentially full RTP for the entire inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R04 Equipment Alignment

a. Inspection Scope

Partial System Walkdowns. The inspectors performed partial walkdowns of the following three systems to verify correct system alignment. The inspectors checked for correct valve and electrical power alignments by comparing positions of valves, switches, and breakers to the documents listed in the Attachment. Additionally, the inspectors reviewed the condition report database to verify that equipment alignment problems were being identified and appropriately resolved.

Unit 2 B train safety injection system when the A train was out of service for planned maintenance Unit 2 B train emergency diesel generator (EDG) when the A train EDG was out of service for planned maintenance Unit 1 auxiliary feedwater (AFW) system prior to removal of the turbine driven AFW pump for planned maintenance

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope

Fire Drill Observation. On January 30, inspectors observed a fire drill from the control room. The fire was simulated to be in the low voltage switchyard. On February 13, the inspectors observed the same drill on another group of watchstanders, this time at the low voltage switchyard. The inspectors assessed the adequacy of the fire drill and fire brigade response using licensee procedures 92000-C, Fire Protection Program; 92005-C, Fire Response Procedure; 92030-C, Fire Drill Program; 92916-1, Zone 516 - Low Voltage Switchyard Fire Fighting Preplan; and 17103A-C, Annunciator Response Procedures for the Fire Alarm Computer. The inspectors evaluated the fire brigade

5 performance to verify that they responded to the fire in a timely manner, donned proper protective clothing, used self-contained breathing apparatus, and had the equipment necessary to control and extinguish the fire. The inspectors assessed the adequacy of the fire brigade's fire fighting strategy including entry into the fire area, communications, search and rescue, and equipment usage. Fire Area Tours. The inspectors walked down the following five plant areas to verify the licensee was controlling combustible materials and ignition sources as required by procedures 92015-C, Use, Control, and Storage of Flammable/Combustible Materials, and 92020-C, Control of Ignition Sources. The inspectors assessed the observable condition of fire detection, suppression, and protection systems and reviewed the licensee's fire protection Limiting Condition for Operation log and condition report (CR)database to verify that the corrective actions for degraded equipment were identified and appropriately prioritized. The inspectors also reviewed the licensee's fire protection program to verify the requirements of Updated Final Safety Analysis Report Section 9.5.1, Fire Protection Program, and Appendix 9A, Fire Hazards Analysis, were met.

Documents reviewed are listed in the Attachment.

Unit 1 and Unit 2 control building Level 3 normal air conditioning (AC) room Unit 1 auxiliary building switchgear and mechanical AC heat exchanger rooms Unit 1 containment spray pump room Unit 1 charging pump room Unit 2 auxiliary building switchgear and mechanical AC heat exchanger rooms

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance

a. Inspection Scope

Triennial inspection. The inspectors reviewed inspection records, test results, and other documentation to ensure that heat exchanger (HX) deficiencies that could mask or degrade performance were identified and corrected. The test procedures and records were also reviewed to verify that these were consistent with Generic Letter 89-13 licensee commitments and industry guidelines. The inspectors reviewed documentation associated with the Component Cooling Water and Auxiliary Component Cooling Water, Essential Chill Water, and Residual Heat Removal (RHR) Pump Motor Cooler HXs to assess the health of each. In addition, the inspectors reviewed documentation associated with the Nuclear Service Cooling Water (NSCW) system to assess its capabilities to support these and other risk-significant HXs. Documents reviewed are listed in the Attachment. The inspectors reviewed site and corporate HX program procedures, Maintenance Procedures including testing and cleaning frequencies, design basis documents, CRs, system health reports, Self Assessment Report and conducted interviews with the

6 NSCW system engineer. The inspectors reviewed visual inspection records, flow measurement trends, system walk down inspection results, and eddy current testing procedures. In addition, the inspectors conducted a walk down of the NSCW system and selected HXs to assess general material condition and to identify any degraded conditions of selected components.

CRs were reviewed for potential common cause problems and problems which could affect system performance to confirm that the licensee was entering issues into the corrective action program and initiating appropriate corrective actions.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

a. Inspection Scope

Resident Quarterly Observation. The inspectors observed operator performance on January 23, during licensed operator simulator training described on simulator exercise guide Dynamic Simulator Scenarios V-RQ-SE-08100, V-RQ-SE-08101, and V-RQ-SE-08102. The simulator scenarios covered operator actions resulting from an uncontrolled continuous rod motion, reactor trip, and various instrument failures. Documents reviewed are listed in the Attachment. The inspectors specifically assessed the following areas: Correct use of the abnormal and emergency operating procedures Ability to identify and implement appropriate actions in accordance with the requirements of the technical specifications Clarity and formality of communications in accordance with procedure 1000-C, Conduct of Operations Proper control board manipulations including critical operator actions Quality of supervisory command and control Effectiveness of the post-evaluation critique

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed one equipment problem to evaluate the effectiveness of the licensee's handling of equipment performance problems and to verify the licensee's maintenance efforts met the requirements of 10 CFR 50.65 (the Maintenance Rule) and licensee procedure 50028-C, Engineering Maintenance Rule Implementation. The reviews included adequacy of the licensee's failure characterization, establishment of

7 performance criteria or 50.65(a)(1) performance goals, and adequacy of corrective actions. Other documents reviewed during this inspection included control room logs, system health reports, the maintenance rule database, and maintenance work orders.

Also, the inspectors interviewed system engineers and the maintenance rule coordinator to assess the accuracy of identified performance deficiencies and extent of condition.

CR 2008100319, Unit 2 sequencer B circuit board failure

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation

a. Inspection Scope

The inspectors reviewed the following five work activities to verify plant risk was properly assessed by the licensee prior to conducting the activities. The inspectors reviewed risk assessments and risk management controls implemented for these activities to verify they were completed in accordance with procedure 00354-C, Maintenance Scheduling, and 10 CFR 50.65(a)(4). The inspectors also reviewed the CR database to verify that maintenance risk assessment problems were being identified at the appropriate level, entered into the corrective action program, and appropriately resolved.

Unit 1 NSCW pump number 7 failure 500 KV Switchyard bus removed from service for planned maintenance Planned RHR system outage Turbine Driven AFW (TDAFW) pump removed from service for repairs and planned maintenance Unit 2 operation during the shutdown and cooldown of Unit 1 for planned refueling outage

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following five evaluations to verify they met the requirements of procedure NMP-GM-002, Corrective Action Program, and NMP-GM-002-001, Corrective Action Program Instructions. The scope of this inspection included a review of the technical adequacy of the evaluations, the adequacy of compensatory measures, and the impact on continued plant operation.

CR 2008102694, Unit 1 containment cooler #7 failed to start in high speed CR 2008100319, Unit 2 sequencer B circuit board failure

8 CR 2008100348, Unit 1 NSCW pump number 7 alternate path CR 2008100242, Concrete spalling on Unit 1 A train EDG silencer mounting stand CR 2008103027, White residue on Unit 1 pressurizer heater

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors either observed post-maintenance testing or reviewed the test results for the following four maintenance activities to verify that the testing met the requirements of procedure 29401-C, Work Order Functional Tests, for ensuring equipment operability and functional capability was restored. The inspectors also reviewed the test procedures to verify the acceptance criteria were sufficient to meet the Technical Specification (TS)operability requirements.

Work Order (WO) 10714835, Valve operator testing for 1HV9017B, containment spray pump #2 suction valve WO 10800843, Repairs on NSCW pump number 6 feeder breaker WO 10800491, Unit 1 NSCW transfer pump number 7 repair WO 10709439, Valve operator testing for 1HV8920, Unit 1 train B safety injection pump miniflow isolation valve

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

The inspectors performed the inspection activities described below for the Unit 1 refueling outage that began on March 16. The inspectors confirmed that, when the licensee removed equipment from service, the licensee maintained defense-in-depth commensurate with the outage risk control plan for key safety functions and applicable technical specifications and that configuration changes due to emergent work and unexpected conditions were controlled in accordance with the outage risk control plan.

Documents reviewed are listed in the Attachment.

Prior to the outage, the licensee reviewed the licensee's integrated risk control plan to verify that activities, systems, and/or components which could cause unexpected reactivity changes were identified in the outage risk plan and were controlled Reviewed reactor coolant system pressure, level, and temperature instruments to verify that the instruments provided accurate indication and that allowances were made for instrumentation errors

9 Reviewed the status and configuration of electrical systems to verify that those systems met technical specification requirements and the licensee's outage risk control plan Observed decay heat removal parameters to verify that the system was properly functioning and providing cooling to the core Reviewed system alignments to verify that the flow paths, configurations and alternative means for inventory addition were consistent with the outage risk plan Reviewed selected control room operations to verify that the licensee was controlling reactivity in accordance with the technical specifications Observed the licensee's control of containment penetrations to verify that the requirements of the technical specifications were met Reviewed the licensee's plans for changing plant configuration to verify that technical specifications, license conditions, and other requirements, commitments, and administrative procedure prerequisites were met prior to changing plant configuration

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the following five surveillance test procedures and either observed the testing or reviewed test results to verify that testing was conducted in accordance with the procedures and that the acceptance criteria adequately demonstrated that the equipment was operable. Additionally, the inspectors reviewed the CR database to verify that the licensee had adequately identified and implemented appropriate corrective actions for surveillance test problems.

Surveillance Tests 14980B-2, Diesel Generator 2A Operability Test 24561-1, Unit 1 Train B Reactor Coolant Pump Trip Actuating Device Operational Test In-Service Tests (IST) 14808-1, Unit 1 train A high head safety injection pump Operability Test 14810-1, TDAFW Pump Operability Test 14804-1, Safety Injection Pump Inservice and Response Time Test

b. Findings

No findings of significance were identified.

10 Cornerstone: Emergency Preparedness 1EP6 Drill Evaluation

a. Inspection Scope

The inspectors reviewed the facility activation exercise guide and observed the following emergency response activity to verify the licensee was properly classifying emergency events, making the required notifications, and making appropriate protective action recommendations in accordance with procedures 91001-C, Emergency Classifications, and 91305-C, Protective Action Guidelines.

On February 27, the licensee conducted an emergency preparedness drill involving a loss of coolant accident and a subsequent radioactive release. The technical support center was activated and the site participated in the exercise.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

4OA1 Performance Indicator (PI) Verification

a. Inspection Scope

The inspectors sampled licensee submittals for the listed PIs during the period from January 1, 2007 through December 31, 2007, for Unit 1 and Unit 2. The inspectors verified the licensee's basis in reporting each data element using the PI definitions and guidance contained in procedures 00163-C, NRC Performance Indicator and Monthly Operating Report Preparation and Submittal. Mitigating Systems Cornerstone Unplanned Scrams per 7,000 Critical Hours Scrams with Loss of Normal Heat Removal Unplanned Power Changes per 7,000 Critical Hours The inspectors reviewed portions of the Unit 1 and Unit 2 operator logs, Licensee Event Reports, the monthly operating reports, monthly PI summary reports, the maintenance rule database and NRC inspection reports to verify that the licensee had accurately submitted the PI data.

b. Findings

No findings of significance were identified.

11

4OA2 Identification and Resolution of Problems

a. Inspection Scope

Daily Condition Report Review. As required 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 the licensee's corrective action program. This review was accomplished by either attending daily screening meetings that briefly discussed major CRs, or accessing the licensee's computerized corrective action database and reviewing each CR that was initiated. Focused Review. The inspectors performed a detailed review of the following CRs to verify the full extent of the issue was identified, an appropriate evaluation was performed, and appropriate corrective actions were specified and prioritized. The inspectors evaluated the CR against the licensee's corrective action program as delineated in licensee procedure NMP-GM-002, Corrective Action Program, and 10 CFR 50, Appendix B. Documents reviewed are listed in the Attachment.

CR 2007110447, Evaluation in a potential trend related to emergency light failures

b. Findings and Observations

No findings of significance were identified. When reviewing the potential trend identified in CR 2007110447, the licensee's evaluation compared the emergency light failures noted in October, 2007 to the light failures over the previous two months. Inspectors noted that this was a quarterly surveillance, and to get a good comparison of the failure rates a comparison would have to be made between the failures noted this quarter and the failures noted in the previous quarters. After expanding the comparison window, no adverse trends were identified. Inspectors also noted that this trend evaluation was assigned a severity level of 3 at the Management Review Meeting, but after the evaluators determined that no trend existed, the severity level was reduced to level 4 without providing feedback to the Management Review Meeting. After the inspectors identified the omission to the licensee, the licensee decided to provided feedback to the Management Review Meeting.

4OA3 Event Follow-up

(Closed) Licensee Event Report 05000424/2006-001: Main Reactor Trip due to Loop 3 Main Feed Regulating Valve Control Failure On April 15, 2006, the Unit 1 operating crew observed erratic response of the loop 3 Main Feed Regulating Valve (MFRV) while the plant was operating at full power. The operators were able to stabilize the plant while operating the valve in manual mode. The licensee decided to take the unit to mode 3 to investigate and repair the problem. During the downpower on April 17, the reactor was manually tripped at 33% power when the steam generator number 3 water level was increasing despite the actions of the operators. An investigation by the licensee determined that the failure was due to

12 failures in a transducer in the MFRV control system. The inspectors reviewed the LER, the associated condition report, and subsequent action items. No findings of significance were identified.

4OA5 Other Activities

.1 (Closed) URI 05000424, 425/2007006-003:

Ability To Satisfy TS SR 3.7.5.2

Introduction.

During the component design bases inspection conducted April 23 to May 25, 2007, an unresolved item (URI) was identified related to the failure to evaluate the effects of instrument accuracy, water temperature, or resistance values for installed orifices on surveillance testing of turbine driven auxiliary feedwater pump (TDAFWP) 1/2-1302-P4-001 and motor driven auxiliary feedwater pump (MDAFWP) 1/2-1302-P4-002/003 in accordance with TS Surveillance Requirement (SR) 3.7.5.2. This item was unresolved pending further calculations and analysis by the licensee and NRC review.

This URI was discussed in NRC Report No. 05000424, 425/2007006.

Description.

The inspectors reviewed design calculation X4C1302S12, Auxiliary Feedwater Pump Discharge Line Orifice Sizing, Rev. 1 and calculation X4C1302V04, Auxiliary Feedwater Pumps Technical Specifications Verification, Rev. 5 and identified the following concerns:

The acceptance criterion for the TDAFW pumps was not based on a fluid specific gravity of 1.0. A lower value was used based on an assumed condensate storage tank (CST) temperature of 100 degrees F. This is a non-conservative assumption since the actual CST temperature was lower than this on every test conducted between January 2005 and May 2007.

The acceptance criteria for both TDAFW and MDAFW pumps did not include any allowances for instrument uncertainties. The test results were not corrected for applicable instrument uncertainties. Note: Removed reference to flow resistance values in the Criterion XI NCV.

The inspectors noted that when appropriate instrument uncertainties and actual water temperatures were applied to TS SR 3.7.5.2 test results that the operability of the TDAFWP pumps was called into question. This was due to the fact that the combined effects of instrument uncertainty and correction of water density for actual temperature were larger than the documented available analytical margin for TDAFW pump developed head. This resulted in the licensee performing an operability evaluation on May 11, 2007 as documented in CR2007105436 and RER C070401401.

Analysis.

SNC's failure to consider several factors such as the effects of instrument uncertainty and actual water temperature during TS required surveillance testing was a performance deficiency. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and operability of the auxiliary feedwater (AFW) pumps to perform the intended safety function during a design basis event and the cornerstone attribute of Procedure Quality, i.e. maintenance

13 and testing procedures. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiency did not result in any AFW pumps being inoperable based upon a subsequent review of AFW surveillance testing results when proper instrument uncertainties and water temperatures were applied. This finding was reviewed for cross-cutting aspects and none were identified since the performance deficiencies have existed since initial operation and are not indicative of current licensee performance.

Enforcement.

10 CFR 50, Appendix B, Criterion XI, Test Control states, in part, that test programs shall be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service. Contrary to the above, SNC did not establish adequate test control measures to assure that the MDAFW and TDAFW pumps would be able to perform the intended safety function.

Because this finding is of very low safety significance and was entered into SNC's corrective action program as CR 2007105436, CR 2007105713, CR 2007105870, and CR 2007105895, this violation is being treated as a non-cited violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000424,05000425/2008002-01, Violation of 10CFR50, Appendix B, Criterion XI for Failure To Establish Adequate Test Control Measures For TS SR 3.7.5.2)

.2 (Closed) URI 05000424, 425/2007006-004:

Capability Of Auxiliary Feedwater System To Meet Design And Licensing Requirements

Introduction.

During the component design bases inspection conducted April 23 to May 25, 2007, an unresolved item was identified related to the ability to verify the AFW system was bounded by all operating and design limitations. The inspectors noted that the AFW system had minimal design margin to maintain operability. This item was unresolved pending further calculations and analysis by the licensee and NRC review.

This URI was discussed in NRC Report No. 05000424, 425/2007006.

Description.

The inspectors reviewed design calculation X4C1302S12, Auxiliary Feedwater Pump Discharge Line Orifice Sizing, Rev. 1 and calculation X4C1302V04, Auxiliary Feedwater Pumps Technical Specifications Verification, Rev. 5 and identified the following concerns:

The inspectors reviewed minor design change MDC-00-V1M036, 1A AFW Pump Rotating Element Replacement and Thrust Bearing Change, dated September 11, 2000, and identified that the pump characteristics were changed when the 1A MDAFWP pump rotating element was replaced. The effects of the changes to pump total developed head (TDH) and flow on the AFW system margins were not evaluated.

The accuracy of the AFW system resistance values obtained from plant testing did not evaluate the effect of accuracy of the instrumentation used on the calculation results.

The inspectors reviewed surveillance procedure 28210-C, Main Steam Line Code

14 Safety Valve Setpoint Verification, Revisions 15, 16, 17, and 18 for the steam generator Safety Relief Valves (SRVs) 1PSV3001, 1PSV3011, 1PSV3021, 1PSV3031, 2PSV3001, 2PSV3021, and 2PSV30301 completed from March 2003 through March 2005. This review identified that for these valves the procedure established the as-found acceptance criteria for these valves to be a range 1149 to 1209 psig (1185 +3%, -2%) and the as-left to be a range of 1173 to 1197 psig (1185

+1%, -1%). This allowable band for SRV settings was not applied to the analysis to determine the worst case high steam generator back pressure which could further decrease the available margin for the minimum flow cases.

The inspectors reviewed modification MDD-91-V1M016, Auxiliary Feedwater Rated Speed Change, dated April 4, 1991, and LDCR 2003034, Revise FSAR to Show Values for AFW System Performance Based on Calculation X4C1302S12, dated November 26, 2003. The review identified that in 1991, the TDAFWP speed setting was changed from 4200 to 4230 rpm by modification MDD-91-V1M016, which did not evaluate the effect of the speed increase on the AFW system's margin for the faulted steam generator containment analysis. The licensee has subsequently concluded that the higher TDAFWP speed did not have an adverse effect on the analysis.

Based on the concerns raised during the inspection due to the low margin available for the AFW system, the licensee completely revised Calculation V4C1302V04, Auxiliary Feedwater Pumps Technical Specifications Verification. The calculation was revised to address the changes to 1A MDAFWP TDH as a result of the impeller replacement. The calculation was also revised to take into account worst case high system resistance for the bounding minimum flow cases, worst case high steam generator backpressure of 1235 psig for the bounding minimum flow cases, and worst case low system resistance for the bounding maximum flow cases.

Analysis.

SNC's failure to include the cumulative effects of a new 1A MDAFWP rotating element, accuracy of AFW system resistance values, SRV setpoint tolerances, and TDAFWP speed settings when evaluating the performance of the AFW system was a performance deficiency. This finding is more than minor because it affects the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and operability of the AFW system to perform the intended safety function during a design basis event and the cornerstone attribute of Design Control, i.e. initial design. The inspectors assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) because the deficiencies did not result in the AFW system being inoperable based upon subsequent analysis that showed that the AFW system had sufficient flow performance margin to accommodate pump performance and the increased system flow resistance when applying appropriate resistance values and steam generator backpressures. This finding was reviewed for cross-cutting aspects and none were identified since the performance deficiencies have existed since initial operation and are not indicative of current licensee performance.

Enforcement.

10 CFR 50, Appendix B, Criterion III, Design Control, states, in part, that design control measures be established and implemented to assure that applicable regulatory requirements and the design basis for structures, systems, and components

15 are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, SNC did not include the effects of a new 1A MDAFWP rotating element, accuracy of AFW system resistance values, SRV setpoint tolerances, and TDAFWP speed settings on evaluation of the performance of the AFW system. Because this finding is of very low safety significance and was entered into SNC's corrective action program as CR 2007105979, this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000424,05000425/2008002-02, Capability of Auxiliary Feedwater System to Meet Design and Licensing Requirements)

4OA6 Meetings, Including Exit

.1 Exit Meeting

On April 16, the resident inspectors presented the inspection results to Mr. R. Dedrickson and other members of his staff, who acknowledged the findings. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

.2 Annual Assessment Meeting Summary

On April 24, the Chief, Reactor Projects Branch 2, and the Senior Resident Inspector assigned to the Vogtle Electric Generating Plant met with Southern Nuclear Operating Company representatives to discuss the NRC's Reactor Oversight Process and the NRC's annual safety performance assessment for the period of January 1, 2007 -

December 31, 2007. The major topics addressed were the NRC's assessment program and the results of the licensee's assessment. A listing of attendees and the information presented are available from the NRC's document system (ADAMS) as accession numbers ML081190196 and ML081190202 respectively. ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

R. Brown, Training and Emergency Preparedness Manager

C. Buck, Chemistry Manager

W. Copeland, Performance Analysis Supervisor

R. Dedrickson, Plant Manager

K. Dyar, Security Manager

I. Kochery, Health Physics Manager

J. Robinson, Work Control Superintendent

T. Tynan, Site Vice-President

D. Vineyard, Operations Manager

J. Williams, Site Support Manager

T. Youngblood, Site Engineering Manager

NRC personnel

S. Shaeffer, Chief, Region II Reactor Projects Branch 2

LIST OF ITEMS

OPENED AND CLOSED

Opened and Closed

05000424,

05000425/2008002-01 NCV Failure To Establish Adequate Test Control Measures For TS SR 3.7.5.2 (4OA5.1)

05000424,

05000425/2008002-02 NCV Capability of Auxiliary Feedwater System to Meet Design and Licensing Requirements (4OA5.2)

Closed

05000424/LER-2006-001 LER Main Reactor Trip due to Loop 3 Main Feed Regulating Valve Control Failure (4OA3)

05000424, 425/2007006-003 URI Ability To Satisfy TS SR 3.7.5.2 (4OA5.1)

05000424, 425/2007006-004 URI Capability Of Auxiliary Feedwater System To Meet Design And Licensing Requirements (4OA5.2)

LIST OF DOCUMENTS REVIEWED

Section 1R04: Equipment Alignment Procedures 11105-2, Safety Injection System Alignment

11145-2, Diesel Generator Alignment

11146-2, Diesel Generator Fuel Oil Transfer System Alignment

11610-1, Auxiliary Feedwater System Alignment

2Drawings 2X4DB121, 2X4DB170-2, 1X4DB161

System Health Reports System Health Reports, 3Q2007, Auxiliary Feedwater System (1302A), Diesel Generator System (2403), Safety Injection (1204)

Section 1R05: Fire Protection Procedures 92704-1 Zone 4, Auxiliary Building Wing Area Fire Fighting Preplan 92705-1 Zone 5, Auxiliary Building Level D Fire Fighting Preplan 92719-1 Zone 19, Auxiliary Building Centrifugal Charging Pump Room Fire Fighting Preplan 92720-1 Zone 20, Auxiliary Building

CVCS Pump Room Train A Fire Fighting Preplan 92721-1 Zone 21, Auxiliary Building CVCS NCP Room Fire Fighting Preplan 92723-1 Zone 23, Auxiliary Building Electrical Chase Rooms Fire Fighting Preplan 92723-2 Zone 23, Auxiliary Building Electrical Chase Rooms Fire Fighting Preplan 92835-1 Zone 135, Control Building Level 3 Fire Fighting Preplan

2847-1 Zone 147, Auxiliary Building Level 2 Fire Fighting Preplan 92847-2 Zone 147, Auxiliary Building Level 2 Fire Fighting Preplan 92848-1 Zone 148 Auxiliary Building Level 2 Fire Fighting Preplan 92848-2 Zone 148, Auxiliary Building Level 2 Fire Fighting Preplan 92872-1 Zone 172, Auxiliary Building Level 2 Fire Fighting Preplan 92872-2 Zone 172, Auxiliary Building Level 2 Fire Fighting Preplan 92878-1 Zone 178, Control Building Level 3 Fire Fighting Preplan

2878-2 Zone 178 , Control Building Level 3 Fire Fighting Preplan

2880-1 Zone 180, Control Building Level 3 Fire Fighting Preplan

2880-2 Zone 180, Control Building Level 3 Fire Fighting Preplan Section 1R07 :

Triennial Heat Sink Performance Procedures Procedure No. 83308-C, Rev. 30.3.

Testing of Safety-Related NSCW System Coolers. Procedure No. 83306-C-201, Rev. 7.

CCW and ACCW Heat Exchanger Testing.

Procedure

No. 83306-C-101, Rev. 7.

CCW and ACCW Heat Exchanger Testing.

Procedure

No. 83306-C-202, Rev. 7.

CCW and ACCW Heat Exchanger Testing. Procedure No. 83306-C-204, Rev. 7.

CCW and ACCW Heat Exchanger Testing.

Procedure

83308-C, Rev. 30, Testing of Safety Related NSCW System Coolers

Procedure

83305-C, rev. 7.5, Heat Exchanger Testing/Maintenance Program

Procedure

83306-C, Rev. 7.3, CCW and ACCW Heat Exchanger Testing

Procedure

83309-C, Rev. 6.2, Safety Related Heat Exchanger Inspection

NMP-ES-021, Version 2, Structural Monitoring Program for Maintenance Rule

NMP-ES-009, Version 5, Engineering Programs

NMP-ES-012, Version 2.0, Heat Exchanger Program

NMP-ES-024-701, Version 1, Eddy Current of heat Exchanger Tubing

NMP-GM-003, Version 7.0, Self Assessment Procedure

NMP-GM-004, Version 3.0, Benchmarking

Condition Reports

(CRs)

2002003582,

2002003584,

2002003548,

2008101719,

2008101671,

2008101592,

2008101591,

2008101750,

2008101626

3Miscellaneous Specification X2AP01,Chapter 9. 7, Rev.19. (Civil Code for Construction). 9510-AX4AJ04-155-2.

Parts List for ESF Chiller Unit.

V1-1202-030-H026, Rev. 4.

Pipe Support Drawing for NSCW Anchor Support.

NMAC/EPRI Guidance Manual. "Good Bolting Practices: A Reference Manual for Nuclear Power Plant Maintenance Personnel.

Volume 1: Large Bolt Manual."

Dated 1987.

Vogtle Procedure No. 25009-C, Rev. 18.1.

Concrete Expansion Anchor Installation.

1X6AG02-00016. Safety injection Pump/Motor Manual.

Southern Company Services Calculation Sheet X4C1202V08, Rev. 6.

NSCW/COPATTA-11 Input Data/LOCA Design Case.

Southern Company Services Calculation Sheet X4C2111V01, Version 15.

Control Building HVAC Load.

1-1205-P6-001-M01.

RHR Pump Motor Cooler Trend Data.

1-1205-P6-002-M01.

RHR Pump Motor Cooler Trend Data.

MWO 10203147

1-1203-P4-001-M01.