IR 05000313/2007008

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IR 05000313-07-008 and 05000368-07-008, on 08/25/07 - 09/21/07, Arkansas Nuclear One; Component Design Bases Inspection
ML073120113
Person / Time
Site: Arkansas Nuclear  Entergy icon.png
Issue date: 11/05/2007
From: William Jones
Region 4 Engineering Branch 1
To: Mitchell T
Entergy Operations
References
IR-07-008
Download: ML073120113 (45)
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Text

November 5, 2007Mr. Timothy G. MitchellVice President OperationsArkansas Nuclear One Entergy Operations, Inc.1448 S.R. 333Russellville, AR 72802-0967SUBJECT: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 Information

REGION IV Dockets:50-313; 50-368 License:DPR-51; NPF-6Report:05000313/2007008 and 05000368/2007008Licensee:Entergy Operations, Inc.Facility:Arkansas Nucl ear 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:W illiam 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 or disabled during modifications. The pl ant 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 t

he licensee's probabilisti c 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 bases and that the te sted capabilities met t he 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 mate rial 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)
  • 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
  • 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 ps ig) (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-R Fuses*Information Notice 2007-09: Motor Control Center, Control Power TransformerSizing Concerns*Information Notice 2006-031: Inadequate Fault Interrupting Rating of Breakers

b. Findings

.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 deter mined 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 coul d 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 in jection and cont ainment 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 operat ing procedur e 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 significanc e 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 st ated 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 m onitoring 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 provisi ons 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 calcul ations 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 elec trical 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 functi ons 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 act uator motor tor que 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 t he 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

05000368/2007008-01NCV 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 thru
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