IR 05000498/2007007

IR 05000498-07-007 and 05000499-07-007; on 09/24/2007 - 01/22/2008; South Texas Project, Units 1 and 2; NRC Inspection Procedure 71111.21, Component Design Bases Inspection.

ML080450543
Person / Time
Site:	South Texas
Issue date:	02/13/2008
From:	Russ Bywater Region 4 Engineering Branch 1
To:	Sheppard J South Texas
References
IR-07-007
Download: ML080450543 (35)

Text

UNITED STATES NU CLE AR RE GU LATOR Y C O M M I S S I O N ary 13, 2008

SUBJECT:

SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION, UNITS 1 AND 2 - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000498/2007007 AND 05000499/2007007

Dear Mr. Sheppard:

On November 26, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed onsite portions of a component design bases inspection at your South Texas Project Electric Generating Station, Units 1 and 2. The preliminary results were discussed with you and members of your staff on November 26, 2007. After additional in-office inspection, a telephonic exit was conducted on January 22, 2008. The enclosed report documents our inspection findings.

This 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 team reviewed selected procedures and records, observed activities, and interviewed cognizant plant personnel.

The report documents six NRC identified findings, each involving a violation of NRC requirements. All of the findings were evaluated under the risk significance determination process as having very low safety significance (Green). Because of their very low safety significance and because they are entered into your corrective action program, these violations are being treated as noncited violations, consistent with Section VI.A of the Enforcement Policy.

If you contest the subject or significance of any of these 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, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the South Texas Project Electric Generating Station, Units 1 and 2.

STP Nuclear Operating Company -2-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 Room or from the Publicly Available Records (PARS) component of 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/

Russell L. Bywater, Chief Engineering Branch 1 Division of Reactor Safety Dockets: 50-498; 50-499 Licenses: NPF-76; NPF-80 Enclosures:

NRC Inspection Report 05000498/2007007 and 05000499/2007007 w/Attachment: Supplemental Information cc w/enclosures:

E. D. Halpin Site Vice President STP Nuclear Operating Company South Texas Project Electric Generating Station P.O. Box 289 Wadsworth, TX 77483 Ken Coates Plant General Manager STP Nuclear Operating Company South Texas Project Electric Generating Station P.O. Box 289 Wadsworth, TX 77483 S. M. Head, Manager, Licensing STP Nuclear Operating Company P.O. Box 289, Mail Code: N5014 Wadsworth, TX 77483 C. T. Bowman

STP Nuclear Operating Company -3-General Manager, Oversight STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483 Marilyn Kistler Sr. Staff Specialist, Licensing STP Nuclear Operating Company P.O. Box 289, Mail Code 5014 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 J. J. Nesrsta/R. K. Temple/

E. Alercon/Kevin Pollo City Public Service Board P.O. Box 1771 San Antonio, TX 78296 Jon C. Wood Cox Smith Matthews 112 E. Pecan, Suite 1800 San Antonio, TX 78205 A. H. Gutterman, Esq.

Morgan, Lewis & Bockius 1111 Pennsylvania Avenue NW Washington, DC 20004 Director, Division of Compliance & Inspection Bureau of Radiation Control Texas Department of State Health Services 1100 West 49th Street Austin, TX 78756 Brian Almon Public Utility Commission William B. Travis Building P.O. Box 13326 1701 North Congress Avenue Austin, TX 78701-3326

STP Nuclear Operating Company -4-Environmental and Natural Resources Policy Director P.O. Box 12428 Austin, TX 78711-3189 Judge, Matagorda County Matagorda County Courthouse 1700 Seventh Street Bay City, TX 77414 Anthony Jones, Chief Inspector Texas Department of Licensing and Regulation Boiler Program P.O. Box 12157 Austin, TX 78711 Susan M. Jablonski Office of Permitting, Remediation and Registration Texas Commission on Environmental Quality MC-122, P.O. Box 13087 Austin, TX 78711-3087 Ted Enos 4200 South Hulen Suite 422 Fort Worth, TX 76109 Steve Winn/Christine Jacobs/

Eddy Daniels/Marty Ryan NRC Energy, Inc.

211 Carnegie Center Princeton, NJ 08540 INPO Records Center 700 Galleria Parkway Atlanta, GA 30339-3064 Lisa R. Hammond, Chief Technological Hazards Branch National Preparedness Division FEMA Region VI 800 N. Loop 288 Denton, TX 76209

STP Nuclear Operating Company -5-STP Nuclear Operating Company -6-Electronic distribution by RIV:

Regional Administrator (EEC)

DRP Director (DDC)

DRS Director (RJC1)

DRS Deputy Director (ACC)

Senior Resident Inspector (JLD5)

Branch Chief, DRP/A (CEJ1)

Senior Project Engineer, DRP/A (TRF)

Team Leader, DRP/TSS (CJP)

RITS Coordinator (MSH3)

DRS STA (DAP)

D. Pelton, OEDO RIV Coordinator (DLP1)

ROPreports STP Site Secretary (HLW1)

SUNSI Review Completed: ___Y__ ADAMS: 6 Yes No Initials: ___WSifre___

6 Publicly Available Non-Publicly Available Sensitive 6 Non-Sensitive SRI:EB1 RI:PBC RI:EB1 OE:OB C:EB1 C:PBA C:EB1 WSifre/lm MChambers SMakor GApger RLBywate CEJohnson RLBywater b r

/RA/ /RA/ /RA/ /RA/ /RA/ /RA/ /RA/

1/2/08 12/18/08 1/2/08 1/2/08 2/13/8 2/12/08 2/13/08

STP Nuclear Operating Company -7-OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

SUMMARY OF FINDINGS

IR 05000498/2007007 and 05000499/2007007; September 24, 2007 through January 22, 2008;

South Texas Project Electric Generating Station, Units 1 and 2; NRC Inspection Procedure 71111.21, "Component Design Bases Inspection."

The report covered a 4-week period of onsite inspection and additional in-office inspection performed by six region-based inspectors and two contractors. The inspection identified six Green noncited violations. The significance of most findings is indicated by its color (Green,

White, Yellow, 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 3, dated July 2000.

Cornerstone: Mitigating Systems

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B,

Criterion III, "Design Control," having very low safety significance for the failure to specify in a design calculation allowable relay setpoint tolerances.

Specifically, the licensee failed to specify and verify in the relay setpoint calculations the relay setpoint tolerances used in the calibration test procedures.

The issue was documented in the corrective action program as Condition Record 07-15443.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences.

The failure to verify the effects of relay setpoint tolerances on relay coordination time intervals could have resulted in a loss-of-relay coordination and could lead to either a loss of power to safety-related components or lead to a potential for compromising other equipment on a single fault that the relay was designed to isolate. Using Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green) because the condition did not represent a loss of safety function of a system or a train.

(Section 1R21.b.1)

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B,

Criterion III, "Design Control," having very low safety significance for the failure to include all potential loads in the standby diesel generator fuel oil sizing calculation. Specifically, the licensee did not account for increased standby diesel generator fuel oil usage resulting from the addition of manual electrical loads during the 7-day mission run time. The licensee entered this finding into their corrective action program as Condition Record 07-15592. The licensee subsequently demonstrated that the spent fuel pool cooling pumps would be the only additional manual loads actually used during the 7 days of operation in the bounding design basis scenario and that there were additional conservative assumptions in the sizing calculation to demonstrate sufficient margin.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences.

Using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green)because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. (Section 1R21.b.2)

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B,

Criteria III, "Design Control," of very low safety significance for the failure to translate design basis information into specifications and procedures.

Specifically, a non-conservative system pressure was used as an input to an engineering design calculation for the auxiliary feedwater outside containment isolation valves. This finding has been entered into the licensee's corrective action program as Condition Record 07-15455.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences.

Using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not represent a loss safety function of a system or a train. (Section 1R21.b.3)

Green.

The team identified a noncited violation of Technical Specification Surveillance Requirement 4.8.1.1.2.E.11, having very low safety significance for the licensees failure to adequately perform the technical specification surveillance requirement. Specifically, the licensee failed to verify the loading times of the essential chillers in order to verify the automatic load sequence timer was operable. This issue was entered into the licensees corrective action program as Condition Records 07-14903 and 07-14959.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences.

Using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not represent a loss of safety function of a system or a train.

(Section 1R21.b.4)

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B,

Criterion XI, "Test Control," having very low safety significance for the licensees failure to implement a test program to assure that all installed safety-related molded case circuit breakers will perform satisfactorily in service.

Specifically, the licensee had not adequately exercised or subjected to periodic testing all of the 125V dc molded case circuit breakers since initial plant operation. The licensee entered the finding into their corrective action program as Condition Record 07-15817.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Equipment Performance. It impacts the cornerstone objective of ensuring the availability, reliability, capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A,

Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green) because it did not result in a loss of safety function of a system or a train. (Section 1R21.b.5)

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B,

Criteria III, "Design Control," of very low safety significance for the failure to adequately translate design basis information into specifications and procedures.

Specifically, measurement instrument uncertainties were not included in the determination of minimum allowed high head safety injection pump and low head safety injection pump developed head values used during periodic technical specification surveillance testing. The licensee entered the finding into their corrective action program as Condition Record 07-15752.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences.

Using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not result in a loss of safety function of a system or a train.

(Section 1R21.b.6)

B. Licensee-Identified Findings None.

U.S. NUCLEAR REGULATORY COMMISSION REGION IV Dockets: 05000498, 05000499 Licenses: NPF-76, NPF-80 Report: 05000498/2007007; 05000499/2007007 Licensee: STP Nuclear Operating Company Facility: South Texas Project Electric Generating Station, Units 1 and 2 Location FM 521 - 8 miles west of Wadsworth Wadsworth, Texas 77483 Dates: September 24, 2007 through January 22, 2008k Inspectors: W. Sifre, Senior Reactor Inspector, Engineering Branch 1 M. Chambers, Resident Inspector, Branch C B. Henderson, Reactor Inspector, Engineering Branch 1 S. Makor, Reactor Inspector, Engineering Branch 1 S. Rutenkroger, Reactor Inspector, Engineering Branch 1 G. Apger, Operations Engineer, Operations Branch Contractors: H. Anderson, Mechanical Contractor J. Chiloyan, Electrical Contractor Approved By: Russell L. Bywater, Chief Engineering Branch 1 Division of Reactor Safety

REPORT DETAILS

REACTOR SAFETY

Inspection of component design bases verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected components and operator actions to perform their design bases functions. As plants age, their design bases may be difficult to determine and an important design feature may be altered or disabled during a modification. The plant risk assessment model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems and Barrier Integrity cornerstones for which there are no indicators to measure performance.

In addition to performing the baseline inspection, the team reviewed actions taken by the licensee in response to previously identified significant issues associated with engineering performance.

1R21 Component Design Bases Inspection

The team selected risk-significant components and operator actions for review using information contained in the licensee=s probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum importance value greater than 1E-6.

a. Inspection Scope

To verify that the selected components would function as required, the team reviewed design basis assumptions, calculations, and procedures. In some instances, the team performed independent calculations to verify the appropriateness of the licensee engineers' conclusions. The team also verified that the condition of the components was consistent with the design bases and that the tested capabilities met the required criteria.

The team reviewed maintenance work records, corrective action documents, and industry operating experience information to verify that licensee personnel considered degraded conditions and their impact on the components. For the review of operator actions, the team observed operators during simulator scenarios associated with the selected components, as well as observing simulated actions in the plant.

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modification, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed

performance test results; significant corrective actions; repeated maintenance; 10 CFR 50.65(a)1 status; operable, but degraded, conditions; NRC resident inspector input of problem equipment; system health reports; industry operating experience; and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins.

The components selected for review were:

• 345/138,13.8kV Standby Transformer ST002A

• 13.8kV/4/16 Auxiliary Engineered Safety Feature Transformer E2B

• 4.16kV Engineered Safety Feature BUS E2B

• Standby Diesel Generator 22

• 4.16kV/480 V Load Center Transformer E2B

• 480V Load Center E2B2

• 125V DC Battery and Charger Train B

• Electrical Auxiliary Building HVAC

• 10KVA Inverter E1V 2201

• Auxiliary Feedwater Motor Driven Pump 23

• Auxiliary Feedwater Turbine Driven Pump Td 14

• Auxiliary Feedwater Valve 0019

• Steam Generator Power Operated Relief Valve 2A

• High Pressure Safety Injection Pump 2A

• Low Pressure Safety Injection Pump 2A

• Refueling Water Storage Tank

• Essential Cooling Water Pump 2A

• Essential Chilled Water Pump 2A The selected operator actions were:

• Opening electrical auxiliary building doors and start of smoke purge on loss of ventilation to switchgear rooms.

• Isolation of a faulted steam generator.

• Initiation of reactor coolant system depressurization.

• Tripping of the reactor coolant pumps.

• Diagnosis of a steam generator tube rupture to start appropriate procedures.

• Starting auxiliary feedwater if engineered safety features actuation system fails during a control room fire.

The operating experience issues were:

• NRC Information Notice (IN) 2006-06, "Loss-of-Offsite Power and Station Blackout Are More Probable During Summer Period."

• NRC IN 2007-09, "Equipment Operability Under Degraded Voltage Conditions."

• NRC Generic Letter 89-10, "Consideration of the Results of NRC-Sponsored Tests of Motor-Operated Valves."

• NRC IN 2006-18, "Significant Loss of Safety-Related Electrical Power at Forsmark, Unit 1, in Sweden."

• NRC IN 2007-27, "Recurring Events Involving Emergency Diesel Generator Operability."

b. Findings

b.1. Failure to Specify Setpoint Calibration Limits in Relay Setpoint Calculations

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to specify in a design calculation the allowable relay setpoint tolerances stated in the licensees relay setpoint calibration test procedures. Under postulated electrical fault or overload conditions, the lack of adequate relay coordinating time intervals between relay operating characteristics would lead to spurious tripping and to either a loss of power to safety-related components or lead to a potential for compromising other equipment on a single fault that the relay was designed to isolate.

Description.

During the review of licensees completed protective relay trip setpoint calibration test procedures, relay setting records and relay setting calculations to verify whether the applied relay settings were consistent with the designed basis calculations, the team noted that the acceptance criteria for the allowable values of relay setpoints stated in calibration test Procedures PM EM-2-03000814, WAN 274021 and relay setting sheets were neither specified nor verified in the design basis relay setting Calculation EC-5029, "4.16kV Switchgear Relay Setting." Following discovery, the licensee performed a preliminary evaluation for affected components using the worst-case scenario of relay setpoint tolerances stated on the relay setting records and concluded that the affected components would still perform their required safety functions in the event of an electrical fault. The issue was documented in licensees corrective action program as Condition Record 07-15443.

Analysis.

The licensees failure to specify relay setpoint tolerances and verify the effects on coordination margin in relay setpoint calculations for relays used on 4.16kV emergency safety feature switchgears was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating

events and prevent undesirable consequences. The failure to verify the effects of relay setpoint tolerances on relay coordination time intervals could have resulted in a loss-of-relay coordination and could lead to either a loss of power to safety-related components or lead to a potential for compromising other equipment on a single fault that the relay was designed to isolate. Using Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green) because the condition had not resulted in a loss of safety function of a system or a train. This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, Design Control, requires, in part, that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by performance of a suitable testing program.

Contrary to the above, the licensees design control measures failed to either specify the relay setpoint tolerances or verify the adequacy of the design for safety-related 4160V electrical distribution system to ensure that the trip settings of the protective relays were adequate to ensure selective tripping in the event of a fault. Specifically, the team identified that the licensee failed to specify and verify in the relay setpoint calculations the relay setpoint tolerances used in the calibration test procedures. Because this violation was of very low safety significance and has been entered into the licensees corrective action program as Condition Record 07-15443, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-01, Failure to Specify Setpoint Calibration Limits in Relay Setpoint Calculations.

b.2. Failure to Consider Manual Loads for Fuel Oil Storage Tank Sizing Calculation

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to account for manual electrical loads in determining fuel oil usage during the standby diesel generators seven day mission time for the fuel oil storage tank sizing calculation.

Description.

The Final Safety Analysis Report, Revision 0, stated that the fuel oil storage tanks were sized to have sufficient capacity to provide for continuous operation of the diesel generators for 7 days at their continuous rating, (i.e., 5935 kW). The licensee revised the Updated Final Safety Analysis Report (UFSAR) on December 9, 1992, to replace the loading at the standby diesel generator continuous rating with the engineered safety features load requirements. However, the documented review contained in Unreviewed Safety Question Evaluation 91-0031 and Calculation MC-6256, Sizing of SDG FOST, Revision 0, both discussed including all the non-engineered safety features loads listed in UFSAR, Table 8.3-3, as part of the fuel and storage tank sizing requirement.

In particular, the Unreviewed Safety Question Evaluation 91-0031 stated, This

[including all the listed non-engineered safety features loads] is in accordance with the ANSI N195 Standard which states, If the design includes provision for an operator to supply power to equipment other than the minimum required for the plant condition, such additional load(s) shall be included in the calculation of required fuel oil storage capacity. Regulatory Guide 1.137, Fuel Oil Systems for Standby Diesel Generators, Revision 1, dated October 1979, refers to the requirements described in ANSI N195-1976, "Fuel Oil Systems for Standby Diesel-Generators," to be a method acceptable to the NRC staff for complying with the Commissions regulations regarding diesel fuel oil systems for standby diesel generators and assurance of adequate diesel fuel oil quality. The safety evaluation report originally prepared for South Texas Project Electric Generating Station used ANSI N195 as the standard to evaluate the acceptability of the fuel oil storage tank design and sizing.

Since the UFSAR, as revised, did not discuss the additional manual loads, which must be considered in order to evaluate the fuel oil storage tank sizing, Calculation MC-6256, Revision 0, was ultimately revised in Revision 3, dated October 3, 1996, to remove consideration of all manual loads. Therefore, beginning with that revision the design basis non-conservatively removed consideration of expected actual plant operations with respect to manual loads during the bounding design basis accident analysis.

The team interviewed engineering and operations personnel in order to determine what equipment from UFSAR, Table 8.3-3, would be supplied power other than the minimum required for the plant condition. These interviews revealed a range of possible equipment, which could be utilized since the operations philosophy would be to exceed the minimum required for the plant condition in order to place the plant in as safe a condition as possible. The upper range of potential manually loaded equipment would have resulted in exceeding the minimum technical specification fuel oil volume requirement of 60,500 gallons during the 7-day mission time of the standby diesel generators during the worst-case design basis accident considered. However, in further discussions, licensee personnel balanced the operations philosophy with the 7-day fuel oil requirements considered as part of the design basis event and concluded the spent fuel pool cooling pumps would be the only additional manual loads utilized during the bounding scenario. The fuel oil storage tank sizing calculation included additional conservative assumptions regarding expected pump operation during design basis accident scenarios. For example, the calculation assumed auxiliary feedwater, high head safety injection, and containment spray pumps would be run continuously for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following a large break loss of coolant accident. Therefore, the licensee recalculated the fuel oil storage tank sizing using more realistic assumptions with respect to load profile and determined sufficient fuel oil margin does exist with all design basis conditions considered. This issue was entered into the licensees corrective action program as Condition Record 07-15592.

Analysis.

The team determined that the failure to account for manual electrical loads in determining fuel oil usage during the standby diesel generators 7-day mission time for the fuel oil storage tank sizing calculation was a performance deficiency. The finding

was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Not accounting for the additional manual loads increases the likelihood that the required inventory of fuel oil for a 7-day mission time would not be available.

Using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality.

This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, the licensee had not correctly translated design basis information into the standby diesel generator fuel oil tank sizing analysis. Specifically, the licensee failed to translate the loading and usage associated with additional manual loads, reasonably expected to be utilized during the bounding design basis accident, into Calculation MC-6256, Revision 4. Because this violation was of very low safety significance and has been entered into the licensees corrective action program as Condition Record 07-15592, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-02, Manual Loads not Considered for Fuel Oil Storage Tank Sizing Calculation.

b.3. Failure to Use Correct Design Inputs in Determination of the Weak Link for the Auxiliary Feedwater System Outside Containment Isolation Motor Operated Valves.

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," for the failure to translate design basis information into specifications and procedures. Specifically, the team identified that a non-conservative system pressure was used as an input to the engineering design calculation for the auxiliary feedwater outside containment isolation valves (MC-6204 Document Change Notice MC-145 issued 7/31/1992, Revise Motor-Operator Valve Thrust and Torque Calculation for AF-19, AF-48, AF-65, and AF-85).

Description.

The team identified that the pressure loading calculation in the motor-operated valve weak link analysis for the auxiliary feedwater outside containment isolation valves used a system pressure of 1250 psig. This value was based on the steam generator power-operated relief valves in the main steam systems being normally set at 1225 psig for normal operation and an additional 25 psig was added to the nominal steam generator power-operated relief valve set point to allow for any set point uncertainty.

This did not take into account accident conditions that result in the backpressure from the main steam system being greater than 1250 psig.

In response to the teams questions that the pressure could be greater than 1250 psig, the licensee issued Condition Record 07-15455-4, "Discussion Paper; Re-perform Weak Link Calculation at 1324 psid and 200°F," received October 24, 2007; and Condition Record 07-15455, "Discussion Paper; Weak Link Discussion of Motor-Operated Valves During Normal and Accident Operation," received October 15, 2007. The licensee determined that an increase in steam generator pressure greater than normal operating pressure would occur during certain design bases accident conditions. The appropriate input to the calculation was determined to be a steam generator pressure of 1324 psig, which allows for a 1 percent margin for setting tolerance and 2 percent for pressure drop in the piping connecting the safety valves to the steam generator from the lowest safety valve set point of 1285 psig. With the revised 1324 psig value and the original assumed valve temperature of 200°F the new weak link calculation resulted in two of the eight auxiliary feedwater outside containment isolation valves (one in each unit) having a torque switch setting that exceeded the weak link calculated set point in the close direction. The weak link for these valves is the valve seat. Valve thrust plus system pressure exceeding the valve seat strength could result in thrusting the valve disc into the seat and failure of the valve.

The licensee subsequently provided the following justification for the operability of the valves using the 1324 psid accident pressure. From a review of all accidents that result in an increase in Steam Generator pressures also result in the starting of the auxiliary feedwater pumps. The auxiliary feedwater system water supply has a design temperature range of 32°F to 120°F. Single failure criteria states that one of the auxiliary feedwater pumps may not start, however it is NOT creditable for a pump to not start and to have sufficient back leakage to raise the temperature of the outside containment isolation valve to 200°F at the same time. Therefore, the maximum abnormal temperature is 170°F.

The licensee determined that the weak link calculation at 1324 psid and 170°F results in adequate margin between current thrust settings of all eight auxiliary feedwater outside containment isolation valves and the calculated weak link stresses of the valve seats to assure operability under accident conditions.

Analysis.

The failure to use a conservative design input in the engineering analysis was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green) because it did not represent a loss of safety function of a system or a train. This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," states, in part, that measures shall be established to assure that design basis are correctly translated into specifications and procedures. Contrary to the above, in Calculation MC-145, the licensee did not use a conservative pressure input necessary to prevent damage to auxiliary feedwater outside containment isolation valves during a design basis event. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15455), it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-03, Failure to Use Correct Design Inputs in Determination of the Weak Link for the Auxiliary Feedwater System Outside Containment Isolation Motor Operated Valves.

b.4. Surveillance Procedure Lacked Check for Timing of Chiller Loading on the Bus

Introduction.

The team identified a Green noncited violation of Technical Specification Surveillance Requirement 4.8.1.1.2.E.11, for the licensees failure to adequately perform the technical specification surveillance requirement. Specifically, the licensee failed to verify the loading times of the essential chillers in order to verify the automatic load sequence timer was operable.

Description.

Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 requires Verifying that the automatic load sequence timer is OPERABLE with the first sequenced load verified to be loaded between 1.0 second and 1.6 seconds, and all other load blocks within +/- 10% of its design interval. The team requested to review the strip chart data recorded from the surveillance tests that demonstrated this surveillance requirement had been performed satisfactorily. Licensee personnel recognized, however, that the actual loading times referenced in the surveillance requirement had not been included in the measurements. Procedure 0PSP02-SF-001A, ESF Diesel Sequencer Timing Test Train A, Revision 11 (Trains B and C similar), only tests the time that the sequence timer demands breaker closure and does not measure and/or record the actual load times.

The licensee entered Technical Specification 4.0.3 for all three trains of standby diesel generators for both units, allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to fully perform the surveillances successfully. By reviewing the strip chart recorder data for the last loss-of-offsite power and loss-of-offsite power with emergency safety features actuation testing of the standby diesel generators, the licensee verified Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 was successfully met for Standby Diesel Generators 11, 21, and 22. In the case of Standby Diesel Generators 13 and 23, the recorded information had a time resolution loss due to switching of recording speeds during the test. The licensee performed a risk evaluation to delay the complete performance of the surveillance test until the next scheduled time (the next outage scheduled for Spring 2008). The team reviewed this assessment and agreed with its conclusions since the data that was available fully supported the equipment being able to perform its safety function.

However, for Standby Diesel Generator 12, a review of the strip chart data revealed that Essential Chiller 12B had loaded on the bus at 168 seconds versus the design interval of 270 seconds. This condition had not been discovered in prior surveillance testing because Procedure 0PSP02-SF-001A did not contain instructions to verify the timing of relays outside of the sequence timer itself. The licensee declared Standby Diesel Generator 12 inoperable at 09:45 on October 5, 2007, entering Technical Specification 3.8.1.1, Actions B and D.

The cause of the timing discrepancy was isolated to a 35 second blocking circuit external to the chiller that would not prevent the chiller from performing its design safety function. As such, the safety functions of the sequence timer, the standby diesel generator, and Essential Chiller 12B were not adversely affected by the condition, nor would those safety functions be impacted by starting/loading times of the essential chillers between 65 and 270 seconds. The licensee revised the design documents referencing the loading time of the essential chillers to be between 65 and 270 seconds.

Once completed, the surveillance testing was declared successful, and the licensee declared Standby Diesel Generator 12 operable at 18:25 on October 11, 2007. This issue was entered into the licensees corrective action program as Condition Records 07-14903 and 07-14959.

Analysis.

The team determined that the failure to adequately perform Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 was a performance deficiency.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Not fully performing the required surveillances increases the likelihood that the standby diesel generators and supported equipment would not perform their design safety functions when needed.

Using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green) because the finding did not represent a loss of safety function of the sequence timer, standby diesel generator, or the essential chiller. This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 requires Verifying that the automatic load sequence timer is OPERABLE with the first sequenced load verified to be loaded between 1.0 second and 1.6 seconds, and all other load blocks within +/- 10% of its design interval. Contrary to the above, the licensee failed to verify the actual loading times of the sequenced loads. Specifically, the licensee only verified the time that the sequence timer demands breaker closure and did not perform the verified to be loaded requirement. Because the violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Records 07-14903 and 07-14959, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498;

499/2007007-04, Surveillance Procedure Lacked Check for Timing of Chiller Loading on the Bus.

b.5. Inadequate Test Program for 125V DC Molded Case Circuit Breakers

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to implement a test program to assure that all installed safety-related molded case circuit breakers will perform satisfactorily in service. Specifically, the licensee had not adequately exercised or subjected to periodic testing all of the 125V dc molded case circuit breakers since initial plant operation.

Description.

During the review of surveillance tests for the Auxiliary Feedwater Motor-Operated Valve 0019, the team discovered that the molded case circuit breaker had not been exercised or subjected to testing since the initial plant operation. In addition, further inspection discovered that the majority of 125V dc-fed molded case circuit breakers were also not exercised or subjected to periodic testing since installation in 1986. The types of molded case circuit breakers undergoing any type of preventative testing/maintenance included battery chargers, distribution panels, and inverters since they were infrequently cycled by other maintenance activities. Conversely, the breakers that fed loads to standby diesel generator field flash, reactor trip switchgear, 4.16kV switchgear control power and emergency safety feature load sequencers appeared to have not been tested since it was assumed that they were cycled in other maintenance activities.

The team noted that the licensee performed tests on molded case circuit breakers to satisfy Information Notice IEN 93-64 and ensure that molded case circuit breakers installed remained functional during plant operations. Following the test was an engineering evaluation acknowledging that molded case circuit breakers were subject to potential age-related degradation, which could result in a failure to trip in accordance with the published time-current characteristic curves because of various factors, such as grease hardening. In 2001, the licensee decided that the sample size for the dc-fed loads indicated that limited failures in the test population did not warrant a pre-established test program. Essentially, credit was taken for circuit breakers being cycled as a part of other maintenance programs, but it was realized that these tests performed on breakers, failed to actually cycle the breaker. In fact the handswitch was used to open and close the valve.

Updated Final Safety Analysis Report, Section 8.3.2.1.4, provides for Periodic testing Class 1E dc power system equipment is performed in accordance with Regulatory Guide 1.32 to verify its ability to perform its safety function. Information Notice 93-64, Periodic Testing and Preventative Maintenance of Molded Case Circuit Breakers, stated, "Detecting or assessing degradation could only be accomplished through appropriate periodic testing and monitoring." The team found that the licensees evaluation and approach to the industry experience, design life, potential common mode failures, and component age concerns were not addressed in the test program. The licensee entered this finding into their corrective action program as Condition Record 07-15817.

Analysis.

The team determined that the lack of periodic testing on all of the dc molded case circuit breakers was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Equipment Performance. It impacts the cornerstone objective of ensuring the availability, reliability, capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not result in a loss of safety function of a system or train. This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion XI, "Test Control," stated, 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, the licensee failed to implement a test program to assure all installed safety-related molded case circuit breakers will perform satisfactorily in service. Because this violation was of very low safety significance and has been entered into the licensees corrective action program as Condition Record 07-15817, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-05, Inadequate Test Program for 125V DC Molded Case Circuit Breakers.

b.6. Failure to Incorporate Instrument Uncertainties into Surveillance Requirements for Technical Specification Limiting Condition for Operation 3.5.2 (Surveillance Requirement 4.5.2.f)

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," for the failure to adequately translate design basis information into specifications and procedures. Specifically, measurement instrument uncertainties were not included in the determination of minimum allowed high head safety injection pump and low head safety injection pump developed head values during periodic technical specification surveillance testing.

Description.

Technical Specification Limiting Condition for Operation 3.5.2, Surveillance Requirement 4.5.2.f.1 for the high head safety injection pump, and Surveillance Requirement 4.5.2.f.2 for the low head safety injection pump require:

For the High Head Safety Injection pumps, verification that the pump develops a differential pressure on recirculation flow when tested pursuant to Technical Specification 4.0.5 greater than or equal to 1480 psid.

For the Low Head Safety Injection pumps, verification that the pump develops a differential pressure on recirculation flow when tested pursuant to Technical Specification 4.0.5 greater than or equal to 286 psid.

Upon review of Surveillance Procedures 0PSP03-SI-0001 and 0PSP03-SI-0004, the team identified that the pump developed head acceptance criteria in the procedures did not include consideration of measurement instrument uncertainties and were numerically equal to the technical specification values. As a result, there was no documented assurance that the recorded current and historical surveillance test results would demonstrate pump developed heads above the required minimum technical specification requirements when measurement instrument uncertainties were taken into consideration.

Therefore, the technical specification surveillance test acceptance criteria were non-conservative.

The team reviewed Design Basis Document 5Z529ZB01025, "Technical Specification/

Limiting Conditions for Operation Design Basis Document," Revision 2, and determined that it had erroneously stated for both high-head safety injection pumps and low-head safety injection pumps that This value is a conservative, nominal value and needs no additional instrument uncertainty margin. This value is acceptable for use. This value is only used in this application (Technical Specifications 4.5.2.f.1 and Technical Specification 4.5.2.f.2).

The licensee issued Condition Record 07-15752. The condition record stated that The pump test procedures currently use the technical specification values as the low limit for operability and should be revised. The most recent performance of all safety injection pumps meets the upward adjusted low limits.

Analysis.

The failure to include consideration of measurement instrument uncertainties, in relation to the instrumentation utilized in periodic surveillance tests to measure the pump developed head, into the technical specification surveillance test acceptance criteria was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, the finding screened as having very low safety significance (Green) because it

did not represent a loss of safety system function. This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, states, in part, that measures shall be established to assure that design bases are correctly translated into specifications and procedures. Contrary to the above, the licensee did not conservatively account for the effect of instrument uncertainty in development of acceptance criteria for the technical specification surveillance values for Technical Specification Limiting Condition for Operation 3.5.2. Thus, the minimum allowed high head safety injection and low head safety injection pump developed head had not been definitively demonstrated during surveillance testing to exceed the minimum Technical Specification limiting condition for operation values. Because this violation was of very low safety significance and has been entered into the licensees corrective action program as Condition Record 07-15752, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-06, Failure to Incorporate Instrument Uncertainties into Surveillance Requirements for Technical Specification Limiting Condition for Operation 3.5.2 (Specifically Surveillance Requirement 4.5.2.f).

4OA5 Other Activities

a.1 Unresolved Item Associated with the Effect of Standby Diesel Generator Technical Specification Voltage Variation on Supplied Equipment

Introduction.

The team identified an unresolved item associated with the steady state output voltage supplied by the standby diesel generators is allowed to vary by Technical Specification 3/4.8.1 from 3744 V to 4576 V (+/- 10%) during a loss of offsite power event. Specifically, the licensee has not analyzed for the effect of this full variation.

Description.

The design analysis assumed maximum supplied voltage variations based upon offsite power supplies which were analyzed to vary less than the technical specification allowed steady state variation for the standby diesel generators.

Components throughout the plant would be adversely affected by either an undervoltage or overvoltage condition.

Since this is a very broad issue that encompasses components powered from the standby diesel generator during a design basis event, the licensee will require significant time to evaluate its effects. Although available safety margins will be less, the degree of this effect is not yet known since the effect of the variation varies upon the analyzed parameter and currently analyzed margins vary significantly. The actual safety function of equipment is not expected to be compromised since the standby diesel generators are presently controlled to a tighter band of voltage operation than allowed by technical specifications and review of the surveillance testing of the standby diesel generators confirms this tighter band is currently being maintained.

Once the licensee has evaluated the effect of the allowed steady state voltage variation and determined the degree of safety margin impact throughout the plant, the NRC can complete the inspection of that analysis in order to close this issue. The licensee has documented this issue in Condition Record 07-15554 and the item is unresolved pending the licensees completion of its analysis and NRC review: URI05000498; 499/2007007-07, Effect of Standby Diesel Generator Technical Specification Voltage Variation on Supplied Equipment.

a.2 Unresolved Item Involving Combined Adverse Conditions not considered in Fuel Oil Storage Tank Sizing

Introduction.

The team identified an unresolved item involving accounting for the combined effect of vortexing and standby diesel generator frequency variations in fuel oil storage tank sizing.

Description.

Calculation MC-6256, Sizing of Standby Diesel Generator Fuel Oil Storage Tank, Revision 4, determined a total 7-day fuel oil requirement of 51,500 gallons, comparing this value with a technical specification requirement of 60,500 gallons. However, this calculation did not consider the effects of vortexing or generator frequency variations. Condition Record 97-14434-10 included an evaluation of fuel oil vortexing completed as part of a Review of Safety Related Tanks (other than Refueling Water Storage Tank & Auxiliary Feedwater Storage Tank) for Vortexing Concerns.

Separately, Calculation EC-5100, Standby Generator Transient Response Model, Revision 2, contained an evaluation performed under Condition Record 97-13089-1 in order to Perform Evaluation of Electrical Frequency Variations on Mechanical Fluid Systems.

The vortexing evaluation determined that 13.5 inches of fuel oil volume would be susceptible to excessive air entrainment, representing 4120 gallons of unusable fuel oil with a 7-day fuel oil requirement of 55,360 gallons (referencing Calculation MC-6038, Standby Diesel Generator Fuel Oil Storage Tank Level Setting Calculation. The total required volume would therefore be 59,480 gallons.

The frequency effects evaluation determined that Standby Diesel Generator load would increase by roughly 6% because the majority of load consists of pumps and fans with primarily friction system loads. The evaluation then compared this 6 percent increase in load with the standby diesel generator fuel oil storage tank calculated margin of more than 10 percent.

However, the vortexing evaluation had already effectively reduced the majority of the analyzed margin with a remaining 1020 gallons of fuel oil between 59,480 gallons and the technical specification requirement of 60,500 gallons. Therefore, applying a 6 percent increase in standby diesel generator load in addition to considering vortexing effects would have exceeded the technical specification requirement under those analyzed conditions.

In addition, the most recent fuel oil storage tank sizing calculation determined a 7 day fuel oil requirement of 51,500 gallons. As discussed in the finding Manual Loads not Considered for Fuel Oil Storage Tank Sizing Calculation, this requirement neglected manual loads during the 7 days for which provision would be made to use during a design basis event. A bounding analysis considering the actual anticipated manual loads, in addition to the vortexing reduction and increased load frequency effect, exceeds the minimum technical specification requirement.

The fuel oil storage tank sizing calculation included additional conservative assumptions regarding expected pump operation during design basis accident scenarios. For example, the calculation assumed auxiliary feedwater, high-head safety injection, and containment spray pumps would be run continuously for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following a large break loss-of-coolant accident. The licensee recalculated the fuel oil storage tank sizing using more realistic assumptions with respect to load profile and determined sufficient fuel oil margin does exist with all design basis conditions considered. This issue was entered into the licensees corrective action program as Condition Request 07-14398 and 07-15592.

After further discussions with staff from the NRC Office of Nuclear Reactor Regulation, the team concluded that this issue of failure to account for the combined effect of vortexing and standby diesel generator frequency variation in the fuel oil storage tank sizing would remain open as an unresolved item. Additional NRC staff review was necessary to determine whether the issue was acceptable, whether it was a finding, or whether it constituted a deviation or violation. Pending completion of this review, this item is unresolved: URI 05000498; 499/2007007-08, Combined Adverse Conditions not Considered in Fuel Oil Storage Tank Sizing.

4OA6 Meetings, Including Exit

On October 26, 2007, the team leader presented the preliminary inspection results to Mr. E. Halpin, Site Vice President, and other members of the South Texas Project staff.

After additional offsite and onsite inspection a preliminary exit meeting was conducted on November 26, 2007, with Mr. J. Sheppard, President and Chief Executive officer and other members of the licensees staff. After additional in-office inspection, a telephonic exit was conducted on January 22, 2008. The licensee acknowledged the findings during each meeting. While some proprietary information was reviewed during this inspection, no proprietary information was included in this report.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

C. Bowman, General Manager Oversight

K. Coats, Plant General Manager

R. Engen, Manager, Maintenance Engineering

E. Halpin, Site Vice President

S. Head, Manager, Licensing

K. House, Manager, Design Engineering

B. Jenewein, Manager, Testing/Programs Engineering

R. Lovell, Manager, Industrial Alliances

M. Meier, General manager Station Support

J. Mertink, Manager, Operations

M. Murray, Manager, Systems Engineering

G. Powell, Manager, Site Engineering

D. Rencurrel, Vise President, Engineering

M. Ruvalcaba, Supervisor, Engineering

J. Sheppard, President and Chief Executive Officer

D. Towler, Manager, Quality

NRC personnel

W. Jones, Chief, Engineering Branch 1

J. Dixon, Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

URI05000498; 499/2007007-07 URI Effect of Standby Diesel Generator Technical Specification Voltage Variation on Supplied Equipment URI05000498; 499/2007007-08 URI Combined Adverse Conditions not Considered in Fuel Oil Storage Tank Sizing

Opened and Closed

NCV05000498; 499/2007007-01 NC Failure to Specify Setpoint Calibration Limits in V Relay Setpoint Calculations NCV05000498; 499/2007007-02 NC Manual Loads not Considered for Fuel Oil Storage V Tank Sizing Calculation Attachment

NCV05000498; 499/2007007-03 NC Failure to Use Correct Design Inputs in V Determination of the Weak Link for the Auxiliary Feedwater System Outside Containment Isolation Motor Operated Valves NCV05000498; 499/2007007-04 NC Surveillance Procedure Lacked Check for Timing of V Chiller Loading on the Bus NCV05000498; 499/2007007-05 NC Inadequate Test Program for 125V DC Molded Case V Circuit Breakers NCV05000498; 499/2007007-06 NC Failure to Incorporate Instrument Uncertainties into V Surveillance Requirements for Technical Specification Limiting Condition for Operation 3.5.2 (Specifically Surveillance Requirement 4.5.2.f)