IR 05000313/2007008

Download: ML073120113

Text

November 5, 2007

Mr. Timothy G. MitchellVice President OperationsArkansas Nuclear One Entergy Operations, Inc.1448 S.R. 333Russellville, AR 72802-0967

SUBJECT: ARKANSAS NUCLEAR ONE, UNITS 1 AND 2 - NRC COMPONENT DESIGNBASES INSPECTION REPORT 05000313/2007008 AND 05000368/2007008

Dear Mr. Mitchell:

On September 21, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed acomponent design bases inspection at your Arkansas Nuclear One station. The enclosedreport documents our inspection findings. The findings were discussed on September 21, 2007,with you and other members of your staff. The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license. The team reviewed selected procedures and records, observed activities, and interviewedcognizant plant personnel.Based on the results of this inspection, the NRC identified one finding that was evaluated underthe risk significance determination process. A violation was associated with this finding. Thefinding was found to have very low safety significance (Green) and the violation associated withthis finding is being treated as a noncited violation, consistent with Section VI.A.1 of the NRCEnforcement Policy. If you contest the noncited violation, or the significance of the violation youshould provide a response within 30 days of the date of this inspection report, with the basis foryour denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. NuclearRegulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011;the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC20555-0001; and the NRC resident inspector at the Arkansas Nuclear One station.

Entergy Operation, Inc.-2-In accordance with 10 CFR 2.390 of the NRC's Rules of Practice, a copy of this letterand its enclosure will be available electronically for public inspection in the NRC PublicDocument Room or from the Publicly Available Records (PARS) component of NRC's documentsystem (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/William B. Jones, ChiefEngineering Branch 1Division of Reactor SafetyDockets: 50-313; 50-368Licenses: DPR-51; NPF-6

Enclosure:

NRC Inspection Report 05000313/2007008 and 05000368/2007008

w/Attachment:

Supplemental Informationcc w/

Enclosure:

Senior Vice President Entergy Operations, Inc.P.O. Box 31995Jackson, MS 39286-1995Senior Vice President & Chief Operating OfficerEntergy Operations, Inc.P.O. Box 31995Jackson, MS 39286-1995Vice President, Operations SupportOperations SupportEntergy Operations, Inc.P.O. Box 31995Jackson, MS 39286-1995Vice PresidentOperations SupportEntergy Operations, Inc.P.O. Box 31995Jackson, MS 39286-1995 Entergy Operation, Inc.-3-Site Vice PresidentEntergy Operations, Inc.Arkansas Nuclear One1448 S. R. 333Russellville, AR 72802General Manager Plant OperationsEntergy Operations, Inc.Arkansas Nuclear One1448 S. R. 333Russellville, AR 72802Director, Nuclear Safety AssuranceEntergy Operations, Inc.Arkansas Nuclear One 1448 S. R. 333Russellville, AR 72802Manager, LicensingEntergy Operations, Inc.Arkansas Nuclear One1448 S. R. 333Russellville, AR 72802Director, Nuclear Safety & LicensingEntergy Operations, Inc.1340 Echelon ParkwayJackson, MS 39213-8298Section Chief, Division of HealthRadiation Control SectionArkansas Department of Health and Human Services4815 West Markham Street, Slot 30Little Rock, AR 72205-3867Section Chief, Division of HealthEmergency Management SectionArkansas Department of Health 4815 West Markham Street, Slot 30Little Rock, AR 72205-3867County Judge of Pope CountyPope County Courthouse100 West Main StreetRussellville, AR 72801 Entergy Operation, Inc.-4-Electronic distribution by RIV:Regional Administrator (EEC)DRP Director (ATH)DRS Director (DDC)DRS Deputy Director (RJC1)Senior Resident Inspector (CHY)Branch Chief, DRP/E (JAC)Senior Project Engineer, DRP/E (GDR)Team Leader, DRP/TSS (CJP)RITS Coordinator (MSH3)DRS STA (DAP)D. Pelton, OEDO RIV Coordinator (DLP)ROPreportsANO Site Secretary (VLH)SUNSI Review Completed: _Y_____ADAMS: YesG No Initials: __RAK____ Publicly Available G Non-Publicly Available G Sensitive Non-SensitiveSRI:EB1RI:EB1RI:EB1SOI:OBC:EB1C:PBEC:EB1RAKopriva/lmbJHNadelJ AdamsMMurphyWBJonesJAClarkWBJones

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11/02/0711/05/0711/02/0711/02/0711/05/0711/02/0711/05/07OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

-1-EnclosureU.S. NUCLEAR REGULATORY COMMISSION REGION IV Dockets:50-313; 50-368 License:DPR-51; NPF-6Report:05000313/2007008 and 05000368/2007008Licensee:Entergy Operations, Inc.Facility:Arkansas Nuclear One, Units 1 and 2Location:Junction of Hwy. 64W and Hwy. 333 South Russellville, ArkansasDates:August 25 through September 21, 2007Team Leader:R. Kopriva, Senior Reactor Inspector, Engineering Branch 1Inspectors:J. Nadel, Reactor Inspector, Engineering Branch 1J. Adams, PhD, Reactor Inspector, Engineeering Branch 1M. Murphy, Senior Operations InspectorAccompanyingPersonnel:P. Wagner, Electrical Engineer, Beckman and AssociatesS. Speigelman, Mechanical Engineer, Beckman and AssociatesApproved By:William B. Jones, ChiefEngineering Branch 1Division of Reactor Safety

-2-Enclosure

SUMMARY OF FINDINGS

IR 05000313/2007008 and 05000368/2007008; 08/25/07 - 09/21/07; Arkansas Nuclear One;Component Design Bases Inspection.The report covers an announced inspection by a team of four regional inspectors, and twocontractors. One finding was identified of very low safety significance. The final significance ofmost findings is indicated by their color (Green, White, Yellow, Red) using Inspection ManualChapter 0609, "Significance Determination Process." Findings for which the significancedetermination process does not apply may be Green or be assigned a severity level after NRCmanagement review. The NRC's program for overseeing the safe operation of commercialnuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3,dated July 2000.A.NRC-Identified Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a noncited violation of 10 CFR Part 50,Appendix B, Criterion III, "Design Control," for multiple nonconservative errors ina Unit 2 emergency diesel generator fuel oil consumption calculation. The errorswere a result of illegible reference data, inconsistently applied methodology, andinadequate calculation reviews, some of which reduced the calculated margin tomeeting design bases requirements. The inspectors determined that the failureto establish an adequate design bases emergency diesel generator fuel oilconsumption calculation constituted a performance deficiency and a violation. The licensee entered this into their corrective action program as ConditionReport ANO-2-2007-01325. The inspectors determined that the violation was more than minor because thefuel oil volume required was called into question by the nonconservative errorsidentified by the NRC and the calculation needed to be performed again usingthe appropriate reference data. In accordance with Inspection ManualChapte 0609, "Significance Determination Process," Appendix A, "SignificanceDetermination of Reactor Inspection Findings for At-Power Situations," theinspectors conducted a Phase 1 screening and determined the finding was ofvery low safety significance (Green) because it was a design deficiencyconfirmed not to result in loss-of-operability in accordance with Part 9900,Technical Guidance, Operability Determination Process for Operability andFunctional Assessment. This issue is being treated as a noncited violationconsistent with Section VI.A of the NRC Enforcement Policy: NoncitedViolation 05000368/2007008-001, Non-conservative Errors in Unit 2 Fuel OilConsumption Calculation (Section 1R21.b.1).

B.Licensee-Identified Violations

None-3-Enclosure

REPORT DETAILS

1.REACTOR SAFETYInspection of component design bases verifies the initial design and subsequentmodifications and provides monitoring of the capability of the selected components andoperator actions to perform their design bases functions. As plants age, their designbases may be difficult to determine and important design features may be altered ordisabled during modifications. The plant risk assessment model assumes the capabilityof safety systems and components to perform their intended safety functionsuccessfully. This inspectable area verifies aspects of the Initiating Events, MitigatingSystems and Barrier Integrity cornerstones for which there are no indicators to measureperformance.

1R21 Component Design Bases Inspection (71111.21)The team selected risk-significant components and operator actions for review usinginformation contained in the licensee's probabilistic risk assessment.

In general, thisincluded components and operator actions that had a risk achievement worth factorgreater than 2 or a Birnbaum value greater than 1E-6.

a. Inspection Scope

To verify that the selected components would function as required, the team revieweddesign bases assumptions, calculations, and procedures. In some instances, the teamperformed calculations to independently verify the licensee's conclusions. The teamalso verified that the condition of the components was consistent with the design basesand that the tested capabilities met the required criteria.The team reviewed maintenance work records, corrective action documents, andindustry operating experience records to verify that licensee personnel considereddegraded conditions and their impact on the components. For the review of operatoractions, the team observed operators during simulator scenarios, as well as duringsimulated actions in the plant.The team performed a margin assessment and detailed review of the selectedrisk-significant components to verify that the design bases have been correctlyimplemented and maintained. This design margin assessment considered originaldesign issues, margin reductions because of modifications, and margin reductionsidentified as a result of material condition issues. Equipment reliability issues were alsoconsidered in the selection of components for detailed review. These included itemssuch as failed performance test results; significant corrective actions; repeatedmaintenance; 10 CFR 50.65(a)1 status; operable, but degraded conditions; NRCresident inspector input of problem equipment; system health reports; industry operatingexperience; and licensee problem equipment lists. Consideration was also given to theuniqueness and complexity of the design, operating experience, and the availabledefense in-depth margins.

The inspection procedure requires a review of 15-20 risk-significant and low designmargin components, 3-5 relatively high-risk operator actions, and 4-6 operatingexperience issues. The sample selection for this inspection was 37 components,7 operator actions, and 7 operating experience issues. The components selected for review were:

• T-55, reactor building sump

• BW-1 (CL), borated water storage tank outlet

• CS-285 (CL), P-7A and P-7B Recirculation to T-41B

• FW-68A (CL), P-7A/B common minimum recirculation to T-41B

• P-7A, Emergency Feedwater Pump (K-3 turbine driven)

• P-7B, emergency feedwater pump (motor driven)

• P-34B, 'B' loop decay heat removal pump

• P-34A, 'A' loop decay heat removal pump

• FW-13B (CL), emergency feedwater to 'B' once through steam generator checkvalve*CV-1277 (CL), 'B' decay heat removal loop discharge to Make-Up Pump P-36Csuction*CV-1276 (CL), 'A' decay heat removal loop discharge to Make-Up Pump P-36Asuction*CV-1406 (CL), Reactor Building Sump Line 'B' outlet valve

• CV-1408 (CL), Borated Water Storage Tank T-3 outlet

• CV-3841 (CL), low pressure injection/Decay Heat Pump Bearing Cooler E-50Binlet valve*CV-3840 (CL), low pressure injection/Decay Heat Pump Bearing Cooler E-50Ainlet*Pressure-operated relief valve

• Emergency feedwater initiation

• Fuel oil supply and transfer

• Safety-related condensate storage tank

• Safety-related batteries*Emergency diesel generator fuel oil (chemistry, consumption, etc.)

• Emergency diesel generator fuel oil storage and day tanks (sizing,instrumentation, vortexing, etc.)*Emergency diesel generator fuel oil transfers pumps (net positive suction head,design inputs, etc.)*Quality condensate storage tank design modification, instrumentation)

• Low temperature over pressure valves Unit 2 (low temperature over pressurepressure relief valves, low temperature over pressure motor-operated valve stopvalves, last 2 years of surveillance test results, etc..)*Emergency diesel generator heat exchangers (some performance test results,vendor manuals, licensee's Generic Letter 89-13 commitments) *Reactor coolant pump seal

• Low temperature over pressure relief valves

• Refueling water storage tank

• Refueling water storage tank discharge valve

• Containment sump discharge valve

• High head safety injection pumps

• Low head safety injection pumps

• Residual heat removal heat exchanger

• Containment spray pump

• Safety injection piggyback valve

• Diesel generator fuel oil transfer system The risk significant operator actions included:

• Scenario 2: Engineered Safeguards Actuation System (reactor coolant systempressure stabilizes less than 150 psig) (Unit 1), Course A1SPGROEOPESAS,Revision 1, dated October 23, 2006.*Scenario: Station Blackout (Unit 2), Course A2SPG-RO-SBO, Revision 3, datedJune 30, 2007.

• Scenario: Control Rod Drive Abnormal Operations (Unit 1),Course A1SPGLOR080101, Revision 0, dated July 30, 2007.*Scenario: Degraded Power (Unit 1), Course A1SPGLOR080102, Revision 0,dated July 30, 2007.*Scenario: Emergency Feedwater Initiation and Control (UnannouncedCasualties) (Unit 1), Course A1SPGLOR080103l, Revision 0, dated July 27,2007.*Scenario: Precise Control (Unannounced Casualties) (Unit 2),Course A2SPGLOR080101, Revision 0, dated July 30, 2007.*Scenario: Federal Response Plan 1 (Code Safety Functional Recovery) (Unit 2),Course A2SPGLOR080102, Revision 0, dated July 10, 2007.The operating experience issues were:

• Ultra low sulfur diesel generator fuel oil

• Asiatic clams

• Service water temperature limits

• Information Notice 2006-021: Containment Sump Voiding

• Information Notice 2006-005: Possible Defect in Bussmann KWN-R and KTN-RFuses*Information Notice 2007-09: Motor Control Center, Control Power TransformerSizing Concerns*Information Notice 2006-031: Inadequate Fault Interrupting Rating of Breakers

.1 Nonconservative Errors in Unit 2 Fuel Oil Consumption CalculationIntroduction:

The inspectors identified a noncited violation of 10 CFR Part 50Appendix B, Criterion III, "Design Control", for Multiple Non-conservative Errors in a Unit2 Emergency Diesel Generator Fuel Oil Consumption Calculation". Contributing factorsto the errors were illegible reference data, inconsistently applied methodology, andinadequate calculation reviews.

Description:

Emergency diesel generator fuel oil consumption calculations shouldemploy conservative assumptions to show that under worst-case conditions, thereremains enough fuel oil storage capacity on site for the emergency diesel generators tofulfill their safety functions.

Arkansas Nuclear One, Unit 2, Calculation 91-E-0107-04, "Emergency Diesel GeneratorFuel Oil Consumption," used the original diesel vendor factory test data as input tocalculate the highest case consumption rate and the required fuel oil storage margin tomeeting all of the safety analysis run time assumptions. The safety analyses require run times based on 50, 100, and 110 percent diesel loadratings. The vendor "Final Acceptance" test that is used as input to the calculation ranthe diesel at all three load ratings and recorded information at each one, including fueloil consumption rate. The calculation methodology took the consumption rates recordedduring the test and corrected them to worst-case scenarios to get a calculatedworst-case consumption at each diesel load rating. These values where then used tocalculate maximum run times at different ratings, given the various fuel oil volumesavailable in different safety analysis assumptions.

The vendor test data sheet was recorded by hand on the day of the test, October 26,1979. The licensee's official copy of this record is from an electronic system to whichthe original was either scanned or microfilmed. This test document carried a warningcover sheet indicating that parts of the document have been determined to be illegibleand no better copy exists in the official records. The main inputs to the calculation from the vendor data sheet are the recorded fuelconsumption rates in lb/hr at each load rating. The vendor performed multiple runs ofthe diesel at each of the three load ratings. The calculation author attempted to pick thehighest recorded consumption rate for each load rating as the calculation input, whichwas the most conservative approach. Upon close inspection of the document, itappears that at 50 percent load rating, 835 lb/hr was the highest recorded value. However, inspectors questioned this value because of the poor quality of the document. The licensee was able to obtain an unofficial record of the test data directly from thevendor. Looking at the unofficial copy, one of the 50 percent load rating runs hadrecorded a consumption of 837 lb/hr, which appeared as 831 lb/hr on the illegible copy. Based on the inspectors review, the 837 lb/hr value is more conservative and shouldhave been used.The calculation employed a correction factor to the consumption numbers taken from thetest data in order to correct for worst-case fuel low heating values. The heating valuerecorded during the test was 19,678 BTU/hr. Arkansas Nuclear One test the fuel oil forlow heating value. The calculation utilized the lowest low heating value the site hadrecorded in the past 5 years at the time of the calculation (1997) and found it to be17,847 BTU/lb. The consumption values, at each load rating from the test data, werethen multiplied by the ratio of these two heating values, which increased the vendorconsumption rates by about 10 percent. The inspectors reviewed the licensee's fuel oil testing procedure and noted that althoughlow heating values were recorded for information only, the procedure does place a limiton the range of acceptable values. If the sample result values fall outside this range, theprocedure requires a condition report be written. According to the procedural limits, thelowest acceptable low heat value is 17,065 BTU/lb. The NRC inspectors challenged theuse of 17,847 BTU/lb as the bounding value in the calculation, given that fuel with a lowheating value of 17,065 BTU/lb can be accepted onsite with no action required.

Applying the lower low heating value to the calculation, this nonconservatism results inthe original consumption numbers from the test data being increased by a 15 percentcorrection factor vice the original 10 percent.This error reduces the calculated diesel run time margins for all scenarios in thecalculation. The most sensitive, however, is the calculation of the 7-day safety analysisrequirement for a single diesel train (one storage tank, one day tank) to operate at50 percent rating after a design basis flood event. The original calculation resulted inonly a 0.7 percent margin to meeting this 7-day requirement (7.05 days of run timeavailable). Once the data sheet transcription error, which only affects the consumptionrate at 50 percent power, was applied along with the proper conservative low heat value;the calculation falls short of the 7-day requirement by almost 4 percent. In addition, the inspectors identified a third nonconservative number in the calculationwhen examining the consumption rate listed at 110 percent load. Instead of employingthe same methods used at 50 and 100 percent load by taking the largest consumptionrate at 110 percent load (1636 lb/hr) from the vendor data sheet, the licenseeinterpolated a smaller value of 1622 lb/hr from the data. Using the number from the datasheet reduces the margin in most of the calculation results, although no results werechallenged in this case. Also of concern, the licensee has not obtained actual data from recent diesel generatortesting for the purposes of confirming that the design bases assumptions remain valid. Engine wear and changes in diesel fuel oil quality over the last 30 years can affect theoperating characteristics of the diesel, including fuel oil consumption rates. When presented with these errors and the questioned ability to meet the safety analysesand design requirements, the licensee began scrutinizing the calculation. Eventually,the licensee found a vendor document from the test runs that called out the19,678 BTU/lb heating value used in the calculation correction factor as a high heatvalue. High heating values account for the heat needed to vaporize the water in the fueloil (heat of vaporization). This heat is not useful to the engine, so typically the heatingvalue of interest is the low heat value, which does not include the heat of vaporization ofwater. However, the use of the high heat value in the calculation, while not technicallycorrect, does add conservatism. In correcting that conservative error, the licensee usedan appropriate low heat value, which reduced the correction factor enough tocompensate for the nonconservative errors identified by the NRC. The licensee wasable to show that they could still meet the 7-day requirement for a design basis flood.

Analysis:

The inspectors determined that the failure to recognize multiple errors in adesign bases emergency diesel generator fuel oil consumption calculation constituted aperformance deficiency and a violation of 10 CFR Part 50, Appendix B, Criterion III. Theviolation was more than minor because it required the fuel oil volume calculations to beperformed again to assure the accident analysis requirements were met. In accordancewith Inspection Manual Chapter 0609, "Significance Determination Process,"Appendix A, "Significance Determination of Reactor Inspection Findings for At-PowerSituations," the inspectors conducted a Phase 1 screening and determined the findingwas of very low safety significance (Green) because it was a design deficiencyconfirmed not to result in loss-of-operability in accordance with Part 9900, Technical Guidance, "Operability Determination Process for Operability and FunctionalAssessment." The licensee entered this finding into the corrective action program asCondition Report ANO-2-2007-01325.Enforcement: Criterion III of Appendix B to 10 CFR Part 50 requires, in part, thatmeasures shall be established for the identification and control of design interfaces andfor coordination among participating design organizations. These measures shallinclude the establishment of procedures among participating design organizations forthe review, approval, release, distribution, and revision of documents involving designinterfaces. The design control measures shall provide for verifying or checking theadequacy of design, such as by the performance of design reviews, by the use ofalternate or simplified calculational methods, or by the performance of a suitable testingprogram.Contrary to the above, as of September 17, 2007, the design control measurestaken were not adequate with respect to Calculation 91-E-0107-04, Revision 2,which contained multiple errors that affected the emergency diesel generator fueloil calculation results, some of which reduced the calculated margin to meetingdesign bases requirements. This violation is of very low safety significance andhas been entered into the licensee's corrective action program as ConditionReport ANO-2-2007-01325, and it is being treated as a noncited violation consistentwith Section VI.A of the NRC Enforcement Policy: NCV 05000368/2007008-001,Nonconservative Errors in Unit 2 Fuel Oil Consumption Calculation.

.2 Refueling Water Tank (Unit 2) and Quality Condensate Storage Tank VortexingIntroduction:

The team reviewed the refueling water tank volume and level setpointcalculations, operating procedures, and the transfer to the reactor containment buildingsump, to determine if sufficient water exists in the tank to support the technicalspecification requirements. The refueling water tank was also reviewed to determine ifsufficient water remains in the refueling water tank following transfer to the containmentsump during a loss-of-coolant accident to prevent drawing air into the safety injectionand containment spray pumps. The time for automatic transfer was reviewed by theteam to determine if adequate water remained in the take to allow for "swap-over" to thecontainment sump. The team also reviewed the quality condensate storage tank volume requirement andthe tank level during the transfer from the quality condensate storage tank to the servicewater system by the emergency feedwater system. The team reviewed the calculationsfor Units 1 and 2 and operating procedures for quality condensate storage tank transferto determine if adequate water exists during the transfer operation. The team reviewedthe calculation of required water level in the tank to facilitate the 30 minute transfer fromthe quality condensate storage tank to the service water system.

Description:

The automatic refueling water tank recirculation actuation system setpointis documented in Calculation 93-EQ-2001-03. This includes the instrument uncertaintyconsiderations, as well as development of the recirculation actuation system setpoint,primarily to show how safety analysis volumes are achieved. For the refueling watertank, the team found that although sufficient water exists to support the technicalspecification requirements, Arkansas Nuclear One relies on a vortex suppressor toassure that there will be no air drawn into the safety injection and containment spraypumps prior to swap over. The refueling water tank level that swap over occurs wasestablished based on the use of a vortex suppressor. However, the vortex suppressordoes not have either analytical or test data to support its use. A corrective actiondocument was issued by Arkansas Nuclear One to evaluate the design and determine ifadequate water level exists. Arkansas Nuclear One prepared an operability evaluationand determined that the plant remains operable in the current condition. The quality condensate storage tank level setpoint was established to provide a30 minute supply of water to facilitate the emergency operating procedure mandatedmanual transfer of the suction of the emergency feedwater pumps to its alternate supply. The quality condensate storage tank design uses a vortex suppressor to assure that thewater level during the transfer to the service water system is adequate. The swap overlevel was established based on the use of a vortex suppressor. Although the setpointlevel is supported by Calculations ANO-1: CALC-90-E-0116-07 Setpoint T.6 andANO-2: CALC-90-E-0116-01 Setpoint T.18, the vortex suppressor does not have eitheranalytical or test data to support its use. Arkansas Nuclear One issued a correctiveaction to analyze both vortex suppressors. The NRC will evaluate the adequacy of therefueling water tank and quality condensate storage tank transfer setpoints when thelicensee has completed testing and analysis of the vortex suppressers. (URI 05000313;368/2007008-02; ).Analysis: The NRC will complete a significance determination, if warranted, whenclosing out the unresolved item.Enforcement: The NRC will consider enforcement, if necessary, when closing out theunresolved item..3Water Storage Tanks

a. Inspection Scope

The team evaluated the instrumentation associated with various water storage tanks toensure there was adequate capability (volume, chemical concentration and temperature)to properly fulfill the design provisions stated in the Updated Final Safety AnalysisReport and other design commitments for each of the evaluated tanks. The tanks thatwere reviewed included the Unit 1 borated water storage tank, the Unit 2 refueling watertank and the common qualified condensate water storage tank. The team reviewed thecalculations for the instrumentation utilized for level and temperature monitoring relatedto these storage tanks in order to verify that uncertainties and scaling properties hadbeen properly incorporated into the indication and control circuitry devices. The teamalso reviewed testing and calibration procedures, including the results of recent testing,for the above tanks to evaluate the performance of the reviewed instrumentation.

b. Findings

No significant findings were identified.

.4 Borated Water Storage Tank Temperature Instrumentation

a. Inspection Scope

During the evaluation of the indication and control systems related to the borated waterstorage tank, the team noted that the last calibration of the temperature instrumentationhad been performed on February 12, 2003. Because this temperature instrumentation isused to verify Unit 1 Technical Specification Surveillance Requirement 3.5.4.1, the teamquestioned the appropriateness of the long interval since its last calibration. Thelicensee investigated the calibration history for the borated water storage tanktemperature Transmitter TT-1413, and found that the instrument had been removed fromthe routine, 3-year interval, testing and calibration program on July 20, 2006. Thelicensee initiated Condition Report ANO-1-2007-02041 to evaluate the cause of theinstrument being removed from the routine testing and calibration program and toimplement corrective actions.The team verified that the borated water storage tank level instrumentation and that thecomparable Unit 2 refueling water tank level and temperature instrumentation continuedto be routinely tested and calibrated. A calibration check of the borated water storagetank temperature transmitter and instrumentation string was completed in accordancewith Repetitive Maintenance Task No. 10638 on September 19, 2007. Theinstrumentation was found to be within acceptable limits.

b. Findings

No significant findings were identified..5Station Batteries

a. Inspection Scope

The team evaluated Units 1 and 2 safety-related station batteries to ensure there wasadequate capability to fulfill the design provisions stated in the Updated Final SafetyAnalysis Reports and other design commitments. The team reviewed the dc voltagerequirement calculations for both units to verify that sufficient voltage would be availableat the terminals of selected loads to ensure their proper operation. The team reviewed the testing procedures and the results of recently completedbattery capacity tests to verify that the batteries could perform the safety functionsdescribed in the Updated Final Safety Analysis Report. The team questionedseveral portions of the procedures that were implemented on both Units 1 and 2 toverify that the methodology was in accordance with manufacturer recommendationsand IEEE-450 guidance. The team also reviewed selected condition reports that hadbeen initiated for problems identified during testing of the station batteries to verify thatappropriate actions had been implemented.

The team reviewed the licensee's studies and calculations pertaining to the ability ofeach unit to cope with a station blackout. The team verified that the licensee'scalculations concluded that adequate voltage would be available to perform suchfunctions as 4160V circuit breaker actuations at the end of the station blackout copingperiod. As part of the station blackout reviews, the team reviewed the offsite powersupply system to ensure that an instability on one of the systems (500 or 161 kV) wouldnot result in the loss of the other system. The team also inspected the battery systemsfor both portions of the switchyard to verify that proper maintenance was beingconducted.

b. Findings

No significant findings were identified..6Emergency Diesel Generator Field Flashing

a. Inspection Scope

During a walkdown of the electrical components installed in the facility, the teamquestioned the function of the small 24V batteries installed in the Unit 2 emergencydiesel generator rooms. The team was informed that the batteries (two 12V batteriesconnected in series) had been installed to provide an emergency source of power forfield flashing. The team reviewed the information the licensee had received from theUnit 2 emergency diesel generator vendor stating that 12V would be adequate for fieldflashing of the Unit 2 emergency diesel generator s. The team also questioned the field flash requirements for the Unit 1 emergency dieselgenerator. The team was informed that a test had been conducted that verified residualmagnetism was adequate for field flashing of the Unit 1 emergency diesel generatorsand, therefore, no separate source of power was required. The team reviewed thetemporary test procedure and noted that generator voltage levels were achieved withoutapplying a source of field flashing current.

b. Findings

No findings of significance were identified..7Motor-Operated Valve

a. Inspection Scope

The team selected motor-operated valves installed in both units to determine if thevalves and their actuators could properly fulfill the design functions discussed in theUpdated Final Safety Analysis Report and other design documents. The team selectedthe containment sump outlet valves from both units, the Unit 1 borated water storagetank outlet valves and the Unit 2 refueling water tank outlet valves.

As part of the evaluation of the motor-operated valves, the team reviewed the degradedvoltage analysis and related calculations to ensure that degraded voltage levels hadbeen utilized in the determination of actuator motor torque capabilities. The team reviewed the schematic diagrams for the selected motor-operated valves toverify that the required actuation and interlock signals had been appropriatelyincorporated in the circuitry. The team also evaluated the circuitry related to the motorthermal overload protective feature to verify that the feature was in accordance with thelicensee's commitments to Regulatory Guide 1.106, "Thermal Overload Protection forElectric Motors on Motor-Operated Valves." Regulatory Position C.1 of RegulatoryGuide 1.106 states that the thermal overload feature should be bypassed duringaccident conditions; Regulatory Position C.2 of the Regulatory Guide states that the tripsetpoint of the thermal overload protection devices should be established with alluncertainties resolved in favor of completing the safety-related action. The teamdetermined that Arkansas Nuclear One, Unit 1, was committed to RegulatoryPosition C.2 of Regulatory Guide 1.106 and that Unit 2 was committed to RegulatoryPosition C.1 of the Regulatory Guide. The team verified that the schematic diagrams for the selected Unit 2 motor-operatedvalves indicated that the thermal overload protection feature was bypassed underaccident conditions. The team reviewed the licensee's standards for sizing motor-operated valve thermal overload heaters and noted that the standard referenced theprovisions of Regulatory Position C.2 of Regulatory Guide 1.106 as one of theconsiderations in the motor-operated valve heater selection process. The teamevaluated the thermal overload heater size installed in three of the selected Unit 1motor-operated valves and verified that the installed heater actuation setpoints werehigher than the motor amperage recorded during valve testing.

b. Findings

No findings of significance were identified..8Borated Water Storage Tank Volume (Unit 1)

a. Inspection Scope

The team reviewed the borated water storage tank level setpoint calculations andoperating procedures for the transfer of flow from the borated water storage tank to thereactor containment building sump to determine if sufficient water exists in the tank tosupport the technical specification requirements. The tank was also reviewed todetermine if sufficient water is in the tank following transfer to prevent drawing air intothe safety injection and containment spray pumps. The manual transfer time wasreviewed by the team to assure that the operator could meet the swap-over timerequirements.

b. Findings

No findings of significance were identified.

.9 Borated Water Storage Tank/Refueling Water Tank Discharge Valves (Units 1 and 2):

a. Inspection Scope

The team reviewed the borated water storage tank and refueling water tank dischargevalves to evaluate if the valves will open in accordance within the time requirements andclose during the transfer from the tanks to the containment sump. The team performedthe following activities: reviewed corrective actions, interviewed the system engineersfor Units 1 and 2 to determine if the valves had operating or maintenance issues, andinterviewed the motor-operated valve engineer to determine if the valves had adequatedesign margin. The valve calculations, corrective actions and inservice test results werereviewed.

b. Findings

No findings of significance were identified..10Unit 1 and 2 Containment Sump Discharge Valves

a. Inspection Scope

The team reviewed the containment sump inboard and outboard discharge valves toassure that the valves would function during a emergency core cooling system event. The team reviewed past corrective actions, valve design calculations and test resultsand interviewed the systems engineers for Units 1 and 2 to determine if the valves hadoperating or maintenance issues. Particular focus was provided by the team forcorrective actions for closing of the Unit 2 valves and bypass leakage. The teamreviewed the assumptions for the control room and off-site dose analysis that was usedin the corrective action.

b. Findings

There were no findings of significance identified.11High Pressure Safety Injection Pumps (Unit 2)

a. Inspection Scope

The team reviewed high pressure safety injection pump calculations to assure that flowand net positive suction head margin exists during emergency operations. In addition,the team conducted an interview with the high pressure safety injection system engineerto determine if there are significant maintenance or operational issues that should bereviewed by the team. The team reviewed past seal leakage and bypass flow issuesthat have been corrected. The team also reviewed the Arkansas Nuclear One correctiveactions for regaining net positive suction head margin because of excessive pump flowthat resulted from rotor replacement.

b. Findings

There were no findings of significance identified..12Low Pressure Decay Heat Removal System Piping (Unit 1)

a. Inspection Scope

The team reviewed the section of piping between the decay heat removal block valvesand the inlet check valves to the reactor coolant system. The team determined that thepiping is vented on a regular basis to assure that pressure, because of check valveweeping, is maintained at a low level to assure the function of the check valve. Theteam reviewed the piping valve and pipe rating to assure that the components wouldfunction at maximum design pressure.

b. Findings

There were no findings of significance identified4OTHER ACTIVITIES 4OA6Meetings, Including ExitOn September 21, 2007, the team leader presented the preliminary inspection results toMr. Tim Mitchell, Site Vice-President, and other members of the licensee's staff. Thelicensee acknowledged the findings during each meeting. While some proprietaryinformation was reviewed during this inspection, no proprietary information was includedin this report.Attachments: 1.

Supplemental Information 2. Initial Information Request A1-1Attachment 1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

D. Bentley, Design Engineering Supervisor

B. Berryman, Manager, Operations Unit 1

E. Blackard, Supervisor, Engineering Programs

C. Bregar, Director, Nuclear Safety Assurance

R. Buser, Electrical Design Engineer

J. Browning, Manager, Operations Unit 2

W. Cottingham, EIC Design Engineer

G. Dobbs, Electrical Design Engineering Supervisor

J. Dubbs, Electrical, Instrument and Controls Design Engineeering Supervisor

D. Edgell, Supervisor, System Engineering

J. Hotz, Electrical Design Engineer

D. James, Manager, Licensing

D. McAfee, Electrical Design Engineer

B. Miller, Battery System Engineer

J. Miller, Jr., Manager, System Engineering

T. Mitchell, Vice President, Operations

C. Reasoner, Manager, Engineering Programs and Components

R. Scheide, Licensing Specialist

J. Smith, Jr., Project Manager

F. Van Buskirk, Licensing Specialist

P. Williams, Supervisor, System Engineering

M. Wood, Electrical System Engineer

NRC personnel

W. Jones, Branch Chief, Engineering Branch 1

C. Young, Acting Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and

Closed

05000368/FIN-2007008-01NCV Failure to Correctly Analyze the Consumption Rate of theEmergency Diesel Fuel Oil (Section 1R21.b.1)

Opened

05000313;368/2007008-02URIVortex Issue (Section 1R21.b.2)

A1-2Attachment 1

LIST OF DOCUMENTS REVIEWED

Condition ReportsANO-1-2002-01853ANO-1-2004-01817ANO-1-2004-01832ANO-1-2004-01974ANO-1-2005-00623ANO-1-2005-01412ANO-1-2007-00785ANO-1-2007-01501ANO-1-2007-01833ANO-1-2007-01924ANO-1-2007-02041ANO-1-2007-02078ANO-2-2003-01515ANO-2-2003-01757ANO-2-2004-00619ANO-2-2004-01033ANO-2-2004-02090ANO-2-2005-00396ANO-2-2005-00599ANO-2-2005-00746ANO-2-2005-01109ANO-2-2006-00800ANO-2-2007-01214ANO-2-2007-01325ANO-C-2007-01358ANO-C-2007-01461ANO-C-2007-01475ANO-C-2007-01480CalculationsNumberTitleRevision/DatePRA-A1-01-001ANO-1 PSA Level-1 Model 3p0 Summary Report8/9/0690-E-0116-07ANO1EOP Set

Point Basis Document, pages181 thru 193Rev. 3, Attachment 110/11/0690-E-0116-01ANO2EOP Set

Point Document, pages 163 thru182Rev. 12, Attachment 1Dated 01/24/07

PRA-A2-01-003ANO-2 PSA Level-1 Model 4p01 SummaryReportDated 10/19/0692-E-0078-04Unit 2 EFW System Pump PerformanceRequirements292-E-0077-04Unit 1 EFW System Pump PerformanceRequirements182-D-2086B-41Accuracy Analysis for

LT-4204,

LT-4205, 2LT-0727-1, 2LT-0727-2485-E-0002-01ANO-1 Diesel Generator #1 and #2 Load Study1585-S-00002-01ANO-2 Diesel Generator #1 (2K4A) and #2(2K4B) Loading Calculation1591-E-0107-03EDG Day Tank Capacity , 2T-30 A & B091-E-0107-022T-57 A&B Capacity (Emergency Fuel OilStorage Tanks)191-E-0107-04Emergency Diesel Generator Fuel OilConsumption - Unit 2291-E-0107-06EDG Storage Tank T-57A, T57B CapacityEvaluation1

NumberTitleRevision/DateA1-3Attachment 191-E-0107-07ANO-1 EDG Fuel Consumption - Unit 1091-E-0107-08EDG Minimum Fuel Oil Rating and MaximumConsumption - Unit 1091-E-0107-09EDG Minimum Fuel Oil Rating and MaximumConsumption - Unit 2091-E-0107-10ANO 1&2 Common Fuel Limits193-D-6025-01HPSI Pump Room "C" Pondage Evaluation andSeismic Support Qualification for 2LE-2007-2,000-E-0008-01Containment Spray Header Fill Time090-E-0100-04ANO-2 Containment Flood, Maximum andMinimum Levels398-E-0028-R0Decay Heat Analysis082-D-2086-0532-004Volume of CST T-41B Requiring Tornado MissileProtection382-2086-60 Design Calculation for T41B392E0078-03LPSI System Pump Performance Requirements192-E-0077-02ANO-1-HP System Pump PerformanceRequirements, 092-E-0077-03ANO-1 LPSI Pump Performance Requirements092-E-0078-02 Unit 2 HPSI System Pump PerformanceRequirements390-E-0100-03RST Minimum and Maximum Volume4M-3950-3RWT Capacity Requirements097-E-0212-01BWST Draindown Analysis3981015N101-02BWST Vortex Breaker Sizing Calculation098-E-0044-01RWT Drawdown Analysis397-E0008-01BWST Venting Capability0

NumberTitleRevision/DateA1-4Attachment 197-E-0012-01,PSV-1412 FFME Cover Air Gap Calculation 13-E-0058-01HPI NPSH fom the BWST 0V-2CV-5648-10MOV Torque Swithc Setpoint 0V-2CV-5672-10MOV Torque Switch Setpoints

4V-CV-1406-10MOV Torque Switch Setpoints 3V-2CV-5650-10MOV Torque Switch Setpoints 6V-CV-1276-10MOV Torque Switch Setpoints 5V-CV-1407-10MOV Torque Switch Setpoints 882-D-2086B-41Accuracy Analysis for [QCST Instruments] 483-D-1153-01Error and Setpoint Analyses for BWST 492-E-0009-01AC Motor Operated Valve Terminal Voltage 992-E-0021-01Emergency Duty Cycle and Battery SizingCalculation

992-E-0072-10Hydrogen Evolution of Batteries 2D11, 2D12 and2D13 093-EQ-2001-03Loop Error, Setpoint, and Time ResponseAnalysis for PPS RWT Level Trip 093-R-003-01Offsite Power System Voltage Re-Evaluation(May '05)

6 98-E-0044-01RWT Draindown Analysis 301-E-0044-01QCST Level for Required Tech. Spec. Volume 0ER-ANO-2005-02102D12 Battery 2R17 Performance Test Results 0

A1-5Attachment 1Design Bases Documents/ Training ManualsNumberTitleRevision/DateULD-1-SYS-02Makeup and Purification / High PressureSafety Injection System 4ULD-1-SYS-04Decay Heat Removal / Low Pressure InjectionSystem 5STM 1-05System Training Manual, Decay Heat RemovalSystem 14STM-1-04System Training Manual, Primary Make-up andPurification 7ULD-2-SYS-02High Pressure Safety Injection System 4ULD-2-SYS-04Low Pressure Safety Injection and ShutdownCooling System 4STM 2-05System Training Manual, Emergency CoreCoolant System 18DrawingsNumberTitleRevision/DateM-2204P&ID Emergency Feedwater 65M-2202P&ID Lube Oil, Lube Oil Cooling,Electro/Hydraulic Controls and Main Steam 20M-2206P&ID Steam Generator Secondary System 146M-204P&ID Emergency Feedwater Storage 162HBD-124-11Pipe Support Details Fuel Oil Storage Tank 2T-57A & 2T-57B Vent Piping 02HBD-124-2Small Pipe Isometric Emergency Diesel FuelTank 2T-57A Vent 42HBD-124-80Small Pipe Isometric Fuel Oil Storage Tank 2T-57A & 2T-57B Vent Piping 32HBD-89-1Large Pipe Isometric Emergency Diesel Fuel Oilto 2P-16A Transfer Pump Inlet 102HBD-90-1Large Pipe Isometric Emergency Diesel Fuel Oilto 2T-57B to 2P-16B 11

NumberTitleRevision/DateA1-6Attachment 12HBD-90-2Small Pipe Isometric 2P-16B Diesel Fuel OilTransfer Pump Inlet Piping4D901-1R10Emergency Diesel Fuel Tank 13'X20' Shell07/29/70D901-1R4Emergency Diesel Fuel Tank 13'X20' Shell07/29/70D901-2R9Emergency Diesel Fuel Tank 13'X20' Shell07/29/7035720-002-0CBI- General Plan for Condensate StorageTank (T-41B) (QCST),035720-036-0Vortex Breaker for bottom Connection for EFW (QCST)0C-46Field Erected Tank, Borated Water StorageTank Details (Sheet 2 and 3) NM230P&ID Reactor Coolant System Sh 1114M-231 Sh 1P&ID Make-up and Purification System109M-232 Sh 1P&ID Decay Heat Removal System1027-DH-1 Sh 1Large Pipe Isometric, Decay Heat Removal toReactor207-DH-2 Sh 1Large Pipe Isometric Decay Heat PumpDischarge to Reactor167-DH-3 Sh 1 Large Pipe Isometric Decay Heat Removal fromReactor197-DG-4 Sh 1Large Pipe Isometric Decay Heat Removal fromReactor237-DH-5 Sh 1Large Pipe Isometric Decay Heat PumpDischarge97-DH-6 Sh 1Large Pipe Isometric Decay Heat PumpDischarge247-DH-7 Sh 1Large Pipe Isometric Decay Heat PumpDischarge107-DH-8 Sh 1Large Pipe Isometric Decay Heat PumpDischarge137-DH-9 Sh 1Large Pipe Isometric Decay Heat Removal22

NumberTitleRevision/DateA1-7Attachment 17-DH-10 Sh 1Large Pipe Isometric Decay Heat PumpDischarge217-DH-11 Sh 1Large Pipe Isometric Decay Heat PumpDischarge217-DH-12 Sh 1 and 2Large Pipe Isometric Eng.

Safeguards PumpSuction Header207-DH-13 Sh1Large Pipe Isometric Decay Heat Pump SuctionHeaders127-DH-14 Sh 1Large Pipe Isometric Primary Make-up PumpSuction Header67-DH-15 Sh 1Large Pipe Isometric Make-up Pipe Suction97-DH-16 Sh 1Large Pipe Isometric Decay Heat Suction FromReactor to Sump177-DH-17 Sh 1Large Pipe Isometric Decay Heat Suction FromReactor Building Sump817-MU-17 Sh 1Large Pipe Isometric Make-up Pipe Suction917-MU-18 Sh 1Large Pipe Iisometric Make-up Pump Suction1117-MU-19 Sh 1Large Pipe Isometric Make-up Pipe Suction12DH-200 Sh 1Small Pipe Isometric, Decay Heat Pump 34Aand P-34B Recirc. Piping9DH-202 Sh 1Small Pipe Isometric, Relief Valve from ReactorCoolant System7DH-203 Sh 1Small Pipe Isometric, HTRS E 35 A and BDrains and Relief valves7DH-204 Sh 1Small Pipe Isometric Press Relief Valves forDecay Heat Removal Pump317-MU-19Large Pipe Isometricf Make Up Suction12M-2232P&ID Safety Injection System116M-2236P&ID Containment Spray System92

NumberTitleRevision/DateA1-8Attachment 12CCB-12-1 Sh 2Large Pipe Isometric High Pressure SafetyInjection Header Loop 1, Rev 442CCB-70-4 Sh 1Large Pipe Isometric HPSI Header Loop 1 toContainment Penetration92CCB-71-5 Sh 1Large Pipe Isometric From HPSI #2 to ReactorCoolant System92DCB-1-1 Sh 1Large Pipe Isometric HPSI Pump P98A to HPSIHeader # 1152DCB-3-2 Sh 1Large Pipe Isometric HPSI Loop 2202P-89B&C Discharge Piping162DCB-3-1 Sh 1Large Pipe Isometric Header 2DCB-1-2 Sh 1 and2Large Pipe Isometric HPSI Number 2 to

CV-5016-2 and 2CV-5056-2112DCB-3-2 Sh 2Large Pipe Isometric High Pressure SafetyInjection Header Loop 202GCB-3-1 Sh 1 and2Large Pipe Isometric Low Pressure SafetyInjection Pump 2P60A192GCB-32-2 Sh 1Large Pipe Isometric, LPSI Pump Discharge62GCB-3-3 Sh 1Large Pipe Isometric LPSI Pump Discharge122GCB-7-1Large Pipe Isometric LPSI Discharge Header52HCB-24-1 Sh 1Large Pipe Isometric Refueling Water Tank 2T-3 to Cont Spray Pumps52HCB-24-2 Sh 1Large Pipe Isometric, Refueling Water Tank 2T-3 to Cont, Spray Pumps62HCB-26-1 Sh 1Large Pipe Isometric, Containment Spray Pump2P-35A Supply152HCB-27-1 Sh 12Large Pipe Isometric Containment Spray Pump2P-35B, Supply from Control Valve 2CV-5631-2122HCB-13-1 Sh 16Large Pipe Isometric from containment SprayPump16

NumberTitleRevision/DateA1-9Attachment 12HCB-13-1 Sh 2Large Pipe Isometric from Cont. Sump to Cont.Spray Pump2A2001S01ANO 500/161 kV Switchyard20E-1, Sh.1ANO1 Station Single Line Diagram54E-1, Sh.2ANO1 Single Line Diagram 500 kV Switchyard3E-2, Sh.1ANO1 Generator System Meter & RelayDiagram20E-4, Sh.1ANO1 Single Line Diagram Main Supply Meter& Relay26E-5, Sh.1ANO1 4160 Volt System Meter & RelayDiagram25E-16, Sh.1ANO1 Single Line Digaram MCC B55 & B5665E-17, Sh.1ANO1 Red Train Vital AC and DC Single Line &Distribution45E-17, Sh.1AANO1 Green Train Vital AC and DC Single Line& Distribution10E-17, Sh.2ANO1 Single Line & Relay Diagram 125VDCSystem2E-18, Sh.1ANO1 Single Line Digaram MCC B61 & B6274E-19, Sh.2ANO1 Single Line Digaram MCC B57 & B6511E-97, Sh.14160V Engineered Safeguard Bus Feeder ACB24E-97, Sh.1A4160V Engineered Safeguard Bus Feeder ACB8E-99, Sh.14160V Engineered Safeguard Bus A3 Lockout &UV Relays20E-99, Sh.1A4160V Engineered Safeguard Bus A4 Lockout &UV Relays4E-100, Sh.1Diesel Generator DG1 ACB27E-100, Sh.1ADiesel Generator ACB10

NumberTitleRevision/DateA1-10Attachment 1E-102, Sh.1Diesel Generator Engine Control SchematicDiagram29E-102, Sh.1ADiesel Generator Engine Control SchematicDiagram5E-102, Sh.1BDiesel Generator Engine Control SchematicDiagram5E-108, Sh.2Diesel Generator #1 Auxiliary ControlSchematic Diagram19E-138, Sh.1Generator Protection & Lockout Relays26E-182, Sh.1Reactor Building Sump Block Valve

CV-1414 25E-182, Sh.1AReactor Building Sump Block Valve

CV-1415 5E-182, Sh.1BReactor Building Sump to Pump P34A 2E-182, Sh.1CReactor Building Sump to Pump P34B3E-183, Sh. 1Decay Heat Cooler Injection Valve

CV-101Schematic17E-184, Sh.1Borated Water to Pump P34A 10E-184, Sh.1ABorated Water to Pump P34B2E-2001, Sh.1ANO2 Station Single Line Diagram30E-2005, Sh.1ANO2 Single Line Meter & Relay Diagram 4160ESF System29E-2005, Sh.2Single Line Meter & Relay Diagram AACGenerator System1E-2006, Sh.1ANO2 Low Voltage Safety Systems Single LineDiagram42E-2016, Sh.5AAC Generator System 480 Volt MCC 2B1612E-2016, Sh.6AAC Generator System 480 Volt MCC 2B161E-2100, Sh.1Diesel Generator 2DG1 ACB28E-2100, Sh.1ADiesel Generator 2DG1 ACB7

NumberTitleRevision/DateA1-11Attachment 1E-2100, Sh.2Diesel Generator ACBs9E-2102, Sh.2Emergency Diesel Generator 2DG1 Start Circuit#123E-2102, Sh.2AEmergency Diesel Generator 2DG1 Start Circuit#24E-2102, Sh.2BEmergency Diesel Generator 2DG1 EngineProtective Trips4E-2102, Sh.2CEmergency Diesel Generator 2DG1 EngineStop Control4E-2102, Sh.2DEmergency Diesel Generator 2DG1 RelayDevelopment3E-2216, Sh.1Containment Sump Recirculation Valve 2CV-5649-125Engineering ResponsesNumberTitleRevision/DateER-ANO-2002-0528-005HPSI Pump NPSH Margin Improvement0ER-ANO-2000-2804-017ES1.1 Bearing Housing Replacement (Inboardand out-board) 2P-89C- HPI Pump,12/3/01ER-ANO-2003-0528-0000Unit 2 B HPSI Pump Runout Limits0ER-ANO-2005-0228-0002R17 Emergent - Setup MOV 2CV-5648-2 toClose on limit0ER-ANO-2006-0389-000U2 EFW Alignment to QCST Evaluation0ER-ANO-2003-0032-001Volume of CST T 41-B Tornado MissleProtection

2Attachment 1ManualsNumberTitleRevision/Date713.3Instruction Manual for Decay Heat Removal3TD W120.0020Instruction Book for Motor Equipment for B&WCompany (P-P34A Motor)0TDB580-0010Instruction, Operation and Maintenance of ByronJackson Horizontal Double Bearing Pumps(P36A)0TD W120.4160Motor data for West. 69F40038,69F40054,69F40055 and 69F00560TD1075.0210Installation, Operation and Maintenance for LowPressure Safety Injection Pump (2P-60A Pump)0TD1075.0260Ingersoll-Rand High Pressure Safety InjectionPumps (2P-89A Pump)0TD S188.0020Installation, Operation and MaintenanceInstructions, Induction Motors, IntegralHorsepower0TDS188.0010Installation, Operation and MaintenanceInstructions , Siemens-Allis InductionMotors/Generators, Large Frame Vertical0TDS188.0040Installation, Operation and MaintenanceInstructions for Unit 2 High Pressure Safety PumpMotors0TD C173.0020Installation & Operating Instructions for C&DFlooded Cell Standby Battery4TD C173.0030Specifications for C&D LCR Lead CalciumStandby Battery3Operating Experience ReportsNumberTitleRevision/DateIE Bulletin 81-03Flow Blockage of Cooling Water to Safety SystemComponents by Corbicula SP. (Asiatic Clam) andMytilus SF. (Mussel)04/10/81IN 98-22Deficiencies Identified During NRC DesignInspections06/17/98EN 98-906Evaluation of

IN 98-226/16/1998

A1-13Attachment 10CAN058114IE Bulletin 81-03 - Flow Blockage to SafetySystem Components by Corbicula & Mytilus05/22/1981ANO Ultra Low Sulphur Deisel Fuel Oil EvaluationIN 85-03Component Failures Caused By Elevated DCVoltages and

KTN-R FusesIN 2006-005Possible Defect in Bussmann

KWN-R

IN 2006-031Inadequate Fault Interrupting Rating of BreakersIN 2007-009MCC Control Power Transformer Sizing ConcernsOperator Training

ProceduresNumberTitleRevision/DateCourse No.A1SPGROEOPESASScenario 2: ESAS, (RCS pressure stabilizes lessthan 150 psig) (Unit 1) 1 Dated 10/23/06EOP 1202.001Reactor Trip,(Unit 1)Change 30 Dated 1/4/07

EOP 1202.010ESAS,(Unit 1)Change 005-03-0

EOP 1202.002Loss of Subcooling Margin (Unit 1)Change 004-02-0EOP 1203.030Loss of Service Water,(Unit 1)Change 016Pages 10 thru 31 Course No.A2SPG-RO-SBOScenario: Station Blackout (Unit 2) 3 Dated 6/30/07Proc. 2104.029Service Water System Operations 010EOP 2203.022Loss of Service WaterEOP 2203.008Natural Emergencies, Section 4 Loss of LakeDardanelle 13EOP 2202.001Standard Post Trip ActionsChange 009EOP 2202/008Station BlackoutChange 007-00-0EOP 2202.010Standard Attachments, Attachment 25: LoadShedding of Vital Battery LoadsChange 010

A1-14Attachment 1ProceduresNumberTitleRevision/DateEN-DC-195Margin Management 21104.005Reactor Building Spray System Operation 461305.007Reactor Building Isolation and Miscellaneous ValveStroke Test 321104.033Reactor Building Ventilation 621104.029Service Water and Auxiliary Cooling System 641104.004Decay Heat Removal Operating Procedure2104.029Service Water System Operations 601102.015Filling and Draining the Fuel Transfer Canal 241000.113Diesel Fuel Monitoring Program 81015.002Decay Heat Removal and low temperature overpressure System Control 291015.008Unit 2 SDC Control 511402.06624 Month Inspection on Unit One Emergency DieselGenerator Engine02/11/051618.035Sampling and Analyzing Diesel Fuel Oil from DieselFuel Oil Transports82304.134Unit 2 EDG 2K4A Instrumentation Calibration142304.134Unit 2 EDG 2K4A Instrumentation Calibration152304.134Unit 2 EDG 2K4A Instrumentation Calibration165010.015-ATT-5EDG Day Tank Level Instrument Loop ErrorCalculation - Submittal Sheet25120.010Unit 1 & Unit 2 MOV Testing8EN-DC-195Nuclear Management Manual21409.769Unit 1 High Pressure Injection Pump (P-36) RecircPiping Vibration Test),0

A1-15Attachment 11104.004Decay Heat Removal Operating Procedure761104.002Make-up and Purification System Operation702104.040LPSI System Operations Procedure412104-039HPSI System Operation471304.217Unit 1 CST Level Instrument Calibration, Red Train(QCST)011304.218Unit 1 CST Level Instrument Calibration, Green(QCST)001307.063Unit 1 D06 & D07 Battery Surveillance, Supplement 6041307.067Unit 1 D06 & D07 Battery Surveillance, Supplement 2072203.008Natural Emergencies132304.194Unit 2 QCST Level Calibration042304.269Unit 2 Plant Protective System RWT InstrumentCalibration002403.001Unit 2 Battery 2D11 Performance Test102403.002Unit 2 Battery 2D12 Performance Test06EES-12Engineering Standard - Motor Operated ValveElectrical Eval.4EN-LI-102Corrective Action Process10Work orderNumberTitleRevision/DateWO 50144905Perform Quarterly Service Water Pump (P-4A)IAW Procedure 1104.02902/06/07WO 51050625Perform Quarterly Service Water Pump (P-4A)IAW Procedure 1104.02905/03/07WO 51082976Perform Quarterly Service Water Pump (P-4A)IAW Procedure 1104.005/18/07

A1-16Attachment 1WO 51000557Unit 1 Sluice Gate and Service Water BayCleaning and Inspection04/12/06WO 50684335Unit 1 Sluice Gate and Service Water BayCleaning and Inspection01/19/05WO 51001771Unit 1 Sluice Gate and Service Water BayCleaning and Inspection05/16/06WO 50613238Perform The 18M EFW SW Suction PipingFlush/Full Stroke of Service Water07/12/04WO 50985465Perform The 18M EFW SW Suction PipingFlush/Full Stroke of Service Water06/08/05WO 51031177Perform The 18M EFW SW Suction PipingFlush/Full Stroke of Service Water06/16/07WO 51042447Perform the 18 Month Refueling Service WaterFlow Test iaw 1309.01307/31/07WO 50985057Perform the 18 Month Refueling Service WaterFlow Test iaw 1309.01306/08/05WO 50684129Perform the 18 Month Refueling Service WaterFlow Test iaw 1309.01312/11/03WO 51087306Perform Quarterly Reactor Building Spray Pump(P-35B) Test iaw Procedure 1104.005 Sup 508/02/07WO 51028470Perform Quarterly Reactor Building Spray Pump(P-35B) Test iaw Procedure 1104.005 Sup 505/10/07WO 51048156Perform Quarterly Reactor Building Spray Pump(P-35B) Test iaw Procedure 1104.005 Sup 502/21/07WO 51083085Perform Quarterly Reactor Building Isolation &Misc Valve Stroke Test iaw Procedure 1305.007SUP 105/22/07WO 51031066Perform Quarterly Reactor Building Isolation &Misc Valve Stroke Test iaw Procedure 1305.007SUP 106/15/07WO 51051346Perform Quarterly Reactor Building Cooling UnitsVCC-2C/2D Flow Test and Stroke Test iawProcedure 1104.033 Sup 505/04/07

A1-17Attachment 1WO 51084257Perform Quarterly Reactor Building Cooling UnitsVCC-2C/2D Flow Test and Stroke Test iawProcedure 1104.033 Sup 506/21/07WO 51054407Perform Quarterly Reactor Building Cooling UnitsVCC-2A/2B Flow Test and Stroke Test iawProcedure 1104.033 Sup 605/17/07WO 51021516Perform Quarterly Reactor Building Cooling UnitsVCC-2A/2B Flow Test and Stroke Test iawProcedure 1104.033 Sup 602/21/07WO 51087310Perform Quarterly Reactor Building Cooling UnitsVCC-2A/2B Flow Test and Stroke Test iawProcedure 1104.033 Sup 608/2/07WO 51050625Perform Quarterly Service Water Pump (P-4A)Test iaw Procedure 1104.029 Sup 105/31/07WO 51082976Perform Quarterly Service Water Pump (P-4A)Test iaw Procedure 1104.029 Sup 205/18/07WO 51098572Perform Quarterly Service Water Pump (P-4A)Test iaw Procedure 1104.029 Sup 207/08/07WO 51047645Perform Quarterly Service Water Pump (P-4A)Test iaw Procedure 1104.029 Sup 202/21/07WO 51086032Perform Quarterly Service Water Pump (P-4B)Test iaw Procedure 1104.029 Sup 208/02/07WO 51053013Perform Quarterly Service Water Pump (P-4B)Test iaw Procedure 1104.029 Sup 205/16/07WO 51047645Perform Quarterly Service Water Pump (P-4B)Test iaw Procedure 1104.029 Sup 202/21/07WO 51049492Perform Quarterly Service Water Pump (P-4C)Test iaw Procedure 1104.029 Sup 303/21/07WO 51055613Perform Quarterly Service Water Pump (P-4C)Test iaw Procedure 1104.029 Sup 305/18/07WO 51088682Perform Quarterly Service Water Pump (P-4C)Test iaw Procedure 1104.029 Sup 308/22/07WO 51083130Perform the Quarterly Service Water ValveStroke Test Procedure 2104.029, Sup 207/27/07

A1-18Attachment 1WO 51037613Perform the Quarterly Service Water ValveStroke Test Procedure 2104.029, Sup 210/10/06WO 51044341Perform the Quarterly Service Water ValveStroke Test Procedure 2104.029, Sup 201/30/07WO 51050725Perform the Quarterly Service Water ValveStroke Test Procedure 2104.029, Sup 201/20/07WO 51044345[Both Red and Green Train Work] Perform theQuarterly Emergency Feedwater Valve StrokeTesting iaw 2106.06 Supplement 301/30/07WO 51050730[Both Red and Green Train Work] Perform theQuarterly Emergency Feedwater Valve StrokeTesting iaw 2106.06 Supplement 301/21/07WO 51083135[Both Red and Green Train Work] Perform theQuarterly Emergency Feedwater Valve StrokeTesting iaw 2106.06 Supplement 307/27/07WO 51042528[Both Red and Green Train Work] Perform theQuarterly Containment Isolation Valve StrokeTesting iaw 2105.05 Supplement 101/24/07WO 51049470[Both Red and Green Train Work] Perform theQuarterly Containment Isolation Valve StrokeTesting iaw 2105.05 Supplement 102/10/07WO 51055601[Both Red and Green Train Work] Perform theQuarterly Containment Isolation Valve StrokeTesting iaw 2105.05 Supplement 105/24/07WO51088671[Both Red and Green Train Work] Perform theQuarterly Containment Isolation Valve StrokeTesting iaw 2105.05 Supplement 108/01/07WO 50259409Perform the Quarterly, Cold Shutdown ValveTest IAW Procedure 2306.06 Sup 111/21/03WO 51008323Perform the Quarterly, Cold Shutdown ValveTest iaw Procedure 2306.06 Sup 101/30/07WO 51031065Perform the Refueling

Reactor Building Isolation& Misc Valve Stroke Test iaw Procedure1305.007 Sup 206/15/07

A1-19Attachment 1WO 50572552Perform the Refueling

Reactor Building Isolation& Misc Valve Stroke Test iaw Procedure1305.007 Sup 211/25/03WO 51047025Perform Quarterly LPI Pump (P-34A) Test,

CV-1433 Open / Closed Stroke Test,

DH-14A,

DH-13A,

DH-17 Closure Verification, and

CV-1432Fail Safe Closed Test iaw Procedure 1104.004,Sup 1.02/16/07WO 51086030Perform Quarterly LPI Pump (P-34A) Test,

CV-1433 Open / Closed Stroke Test,

DH-14A,

DH-13A,

DH-17 Closure Verification, and

CV-1432Fail Safe Closed Test iaw Procedure 1104.004,Sup 108/02/07WO 51053011Perform Quarterly LPI Pump (P-34A) Test,

CV-1433 Open / Closed Stroke Test,

DH-14A,

DH-13A,

DH-17 Closure Verification, and

CV-1432Fail Safe Closed Test iaw Procedure 1104.004,Sup 105/16/07WO 51050724Perform the Service Water Pump 2P-4AQuarterly Test iaw Procedure 2104.029, Sup. 1A01/04/07WO 51083129Perform the Service Water Pump 2P-4AQuarterly Test iaw Procedure 2104.029, Sup. 1A07/27/07WO 51049603Perform the Service Water Pump 2P-4BQuarterly Test iaw Procedure 2104.029, Sup. 1B02/14/07WO 51055684Perform the Service Water Pump 2P-4BQuarterly Test iaw Procedure 2104.029, Sup. 1B05/30/07WO 51088761Perform the Service Water Pump 2P-4BQuarterly Test iaw Procedure 2104.029, Sup. 1B08/06/07WO 51047718Perform the Service Water Pump 2P-4CQuarterly Test iaw Procedure 2104.029, Sup. 1C02/10/07WO 51053858Perform the Service Water Pump 2P-4CQuarterly Test iaw Procedure 2104.029, Sup. 1C05/24/07WO 51086711Perform the Service Water Pump 2P-4CQuarterly Test iaw Procedure 2104.029, Sup. 1C05/31/07WO 50967821[Both Red and Green Train Work] Perform the 18Month, Integrated Engineering Safeguard Testiaw Procedure 2305.00112/11/04

A1-20Attachment 1WO 51015746[Both Red and Green Train Work] Perform the 18Month, Integrated Engineering Safeguard Testiaw Procedure 2305.00101/30/07WO 50276380[Both Red and Green Train Work] Perform the 18Month, Integrated Engineering Safeguard Testiaw Procedure 2305.00101/28/04WO 50967810[Both Red and Green Train Work] Perform the 18Month, Integrated Engineering Safety FeaturesTest iaw Procedure 2305.00312/11/04WO 51086613Perform Quarterly (P7A) Steam Driven EFWPump Test iaw Procedure 1106.006 Sup. 1208/02/07WO 51048166Perform Quarterly (P7A) Steam Driven EFWPump Test iaw Procedure 1106.006 Sup. 1202/21/07WO 51087924Perform Quarterly (P7B) Motor Driven EFWPump Test iaw Procedure 1106.006 Sup. 1108/02/07WO 51049500Perform Quarterly (P7B) Motor Driven EFWPump Test iaw Procedure 1106.006 Sup. 1103/21/07WO 51201976Emergency Feedwater System Operations08/21/07WO 51086726Emergency Feedwater System Operations05/29/07WO 51050729Emergency Feedwater System Operations02/03/07WO 51083134Emergency Feedwater System Operations04/17/07WO 51098668Emergency Feedwater System Operations07/11/07WO 50265192Calibrate BWST Temperature Transmitter02/11/03WO 50339531Clean, Inspect and Calibrate 2TT-5675 [RWTTemperature]04/21/04WO 50379719Calibrate BWST Temperature Transmitter11/22/99WO 50412604Calibrate BWST Temperature Transmitter09/19/01WO 50976340Clean, Inspect and Calibrate 2TT-5675 [RWTTemperature]09/27/05WO 51020065Clean, Inspect and Calibrate 2TT-5675 [RWTTemperature]03/15/07

A1-21Attachment 1WO 51046780Calibrate BWST Red Train Level Transmitter01/18/07WO 51085186Calibrate BWST Red Train Level Transmitter07/10/07WO 51011020ESF Floor Drain Calibration0WO 25307WO 51013867WO 51052594WO 51056204WO 25465WO 51015227WO 51053302WO 51083127WO 30612WO 51015228WO 51053304WO 51084341WO 30614WO 51032671WO 51053842WO 51085476WO 50966046WO 51033509WO 51055037WO 51086611WO 50968270WO 51049985WO 51055038WO 51088019WO 50968271WO 51050630WO 51055468WO 51097689WO 50986709WO 51051347WO 51056203WO 51097689WO 51013866Miscellaneous DocumentsNumberTitleRevision/DateULD-1-SYS-12Arkansas Nuclear One Upper Level DocumentANO Unit 1 Emergency Feedwater System7STM 1-27Arkansas Nuclear One Unit 1 System TrainingManual Emergency Feedwater System12ULD-2-SYS-12Arkansas Nuclear One Upper Level DocumentANO Unit 2 Emergency Feedwater System9STM 2-19-2Arkansas Nuclear One Unit 2 System TrainingManual Emergency Feedwater and AuxiliaryFeedwater Systems27ULD-1-SYS-10Arkansas Nuclear One Upper Level DocumentANO Unit 1 Service Water System14STM 1-42Arkansas Nuclear One Unit 1 System TrainingManual Service and Auxiliary Cooling Water15ULD-2-SYS-10Arkansas Nuclear One Upper Level DocumentANO Unit 2 Service Water System11ER-ANO-2004-Redundant Source of Makeup Water to

ANO-1

2Attachment 10283-000Spent Fuel PoolULD-1-SYS-9ANO Unit 1 Engineered Safeguards ActuationSystem (ESAS)2Form 1015.015Shift Turnover Checklist - Above 280 F Unit 109/12/07Form 1015.015Shift Turnover Checklist - Above 280 F Unit 109/11/071CNA119903Arkansas Nuclear One, Unit No. 1 RE: completionof the Emergency Cooling Pond Licensing BasisReview (TAC M4948)11/19/99EAR 92-6252K3 Low Pressure Steam Supply Concerns06/17/92Fairbanks Morse Colt Test Report - FinalAcceptance10/29/79Fairbanks Morse Colt Test Report - Full Power10/26/79Fairbanks Morse Colt Test Report - Full PowerRe-run10/27/7910 Year Data Plots - Fuel Oil Sample Results08/13/97 to09/12/07Response to Generic Letter 89-13, "ServiceWater

System Problems Affecting Safety-RelatedEquipment01/26/901-EDG Fuel Tank(57A)API Gravity Fuel Oil Sample Results08/1397 to09/12/071-EDG Fuel Tank(57B)API Gravity Fuel Oil Sample Results08/13/97 to09/12/072-EDG Fuel Tank(2T-57B)API Gravity Fuel Oil Sample Results08/13/97 to09/12/072-EDG Fuel Tank(2T-57A)API Gravity Fuel Oil Sample Results08/13/97 to09/12/072E-63APC ET Test Results2E-20APC ET Test Results2E-64APC ET Test Results

A1-23Attachment 12E20BPC ET Test Results2E63BPC ET Test Results2E64BPC ET Test ResultsANIN-910911-131Conversation

Memorandum09/11/91D 4057-88ASTM Standard Practice for manual Sampling ofPetroleum and Petroleum ProductsD 975 - 81ASTM Standard Specification for Diesel Fuel OilsER-ANO-2006-0389-000U2 EFW Alignment to QCST Evaluation2Job Order #00949732Diesel Oil Storage Tank (T-25) Cleaning06/06/96TD E147.0020Maintenance Manual Turbocharged Engine8TD F010.0090Service Information Letter for Fairbanks MorseModel 38D 8-1/8 x 10 OP12TD A310.0060Information for ITT Standard Heat Exchangers,CPK Series0TD A31O.OO3OOperating Instructions and Parts ListType CP & CPRExchangers0TD F010.0220Instructions for Fairbanks Morse Opposed PistonDiesel Stationary Model 38TD8 1/83TD F010.0020Emergency Diesel Plant Operating Instructions6TD F010.0030Instructions Diesel Stationary Model 38TD-1/87TD E147.0010Operating Manual Stationary Power GeneratingUnits (AB20 Generators) Power Take-Off Units0TD F010.0230Marketing Information Letter for Fairbanks MorseParts and Service Operation0TD Y021.0030Care and Maintenance Instructions YoungRadiator Company Charge Air CoolersIntercoolers and Aftercoolers0

A1-24Attachment 151-5000239-00Interim BWOG Report on HPI/MU NozzleCracking06/05/97A-EP-2005-003ANO MOV Program Status Report2P36A, B and C Acceptable Performance TrendCurves09/07Neenah Foundry Co., Inlet Grate Capacities forGutter Flow and Ponded WaterANO1-NE-006-0003ANO-1 Cycle Safety Analysis Groundrules,

1ANO2-NE-07-0001Arkansas Nuclear One Unit 2 Cycle 20 GroundRules,Transmission/Nuclear Reliability UpgradesPresentation07/29/06IEEE Std. 450 Recommended Practices for Maintenance,Testing and Replacement of Large Lead AcidStorage Batteries 1975 and 2002Editions

A2-1Attachment 2The team provided the following information request in writing to the licensee prior to theinspection.Initial Information RequestComponent Design Basis Inspection (71111.21)Arkansas Nuclear One StationPlease provide the following information in order to support the NRC's component design basisinspection effort at your facility.

If there are problems obtaining any of this information, pleasecall the Team Leader, Ronald Kopriva at (817) 860-8104 to discuss alternate arrangements. We would like to have the information ready when we arrive on site for the "bag-man" portion ofthe inspection on August 6, 2007.We prefer, but it's not required, that the information be provided electronically and in asearchable format, such as Adobe, Word, Word Perfect, or Excel.

Other licensee's have foundthat providing the information on a CD is effective and efficient.1.The risk ranking of components from your site specific probabilistic safety analysissorted by Risk Achievement Worth and by Birnbaum Importance.2.A list of your top 500 cutsets from your probabilistic safety analysis.3.Risk ranking of operator actions from you site specific probabilistic safety analysis sortedby Risk Achievement Worth.

Provide copies of your human reliability worksheets forthese items (you may limit this list to the 100 most risk significant actions).4.If you have an external events or fire probabilistic safety analysis model, provide theinformation requested in Items 1 and 2 for external events and fire.5.Any pre-existing evaluation or list of components and calculations with low designmargins (i.e. pumps closest to the design limit for flow or pressure, diesel generatorsclose to design required output, heat exchangers close to rated design heat removaletc.)6.For the last two years, a list of operating experience evaluations, modifications andcorrective actions sorted by component or system.

A one line, or short, description isacceptable.7.A list of any common-cause failures of components in the last 5 years at your facility.8.A list of Maintenance Rule functions.

9.A list of your Maintenance Rule a(1) components.

10.A list of your current temporary modifications.

11.A current list of "operator work arounds."

2.Piping and instrument drawings for your emergency core cooling systems, emergencydiesel generators and off-site power supplies.

At this time, only the mechanical pipingdrawings are needed for the emergency core cooling systems and the emergency dieselgenerators.

A2-2Attachment 2In addition to the above, if available electronically, please provide a copy of each of thefollowing on

CD.1.Final/Updated Safety Analysis Reports

2.Technical Specifications

3.Design Bases Documents for the emergency core cooling systems (including auxiliaryfeedwater), emergency diesel generators and off-site power supplies4.System descriptions or operator training manuals for the emergency core coolingsystems, emergency diesel generators and off-site power supply systemsThank you for your cooperation in these matters.