IR 05000313/2007008

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
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)

Text

ber 5, 2007

SUBJECT:

ARKANSAS NUCLEAR ONE, UNITS 1 AND 2 - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000313/2007008 AND 05000368/2007008

Dear Mr. Mitchell:

On September 21, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed a component design bases inspection at your Arkansas Nuclear One station. The enclosed report 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 and compliance with the Commission's rules and regulations and with the conditions of your license.

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

Based on the results of this inspection, the NRC identified one finding that was evaluated under the risk significance determination process. A violation was associated with this finding. The finding was found to have very low safety significance (Green) and the violation associated with this finding is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest the noncited violation, or the significance of the violation you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the 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 letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

William B. Jones, Chief Engineering Branch 1 Division of Reactor Safety Dockets: 50-313; 50-368 Licenses: 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-6 Report: 05000313/2007008 and 05000368/2007008 Licensee: Entergy Operations, Inc.

Facility: Arkansas Nuclear One, Units 1 and 2 Location: Junction of Hwy. 64W and Hwy. 333 South Russellville, Arkansas Dates: August 25 through September 21, 2007 Team Leader: R. Kopriva, Senior Reactor Inspector, Engineering Branch 1 Inspectors: J. Nadel, Reactor Inspector, Engineering Branch 1 J. Adams, PhD, Reactor Inspector, Engineeering Branch 1 M. Murphy, Senior Operations Inspector Accompanying P. Wagner, Electrical Engineer, Beckman and Associates Personnel: S. Speigelman, Mechanical Engineer, Beckman and Associates Approved By: William B. Jones, Chief Engineering Branch 1 Division of Reactor Safety-1- 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 two contractors. One finding was identified of very low safety significance. The final significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a noncited violation of 10 CFR Part 50,

Appendix B, Criterion III, "Design Control," for multiple nonconservative errors in a Unit 2 emergency diesel generator fuel oil consumption calculation. The errors were a result of illegible reference data, inconsistently applied methodology, and inadequate calculation reviews, some of which reduced the calculated margin to meeting design bases requirements. The inspectors determined that the failure to establish an adequate design bases emergency diesel generator fuel oil consumption calculation constituted a performance deficiency and a violation.

The licensee entered this into their corrective action program as Condition Report ANO-2-2007-01325.

The inspectors determined that the violation was more than minor because the fuel oil volume required was called into question by the nonconservative errors identified by the NRC and the calculation needed to be performed again using the appropriate reference data. In accordance with Inspection Manual Chapte 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," the inspectors conducted a Phase 1 screening and determined the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss-of-operability in accordance with Part 9900,

Technical Guidance, Operability Determination Process for Operability and Functional Assessment. This issue is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: Noncited Violation 05000368/2007008-001, Non-conservative Errors in Unit 2 Fuel Oil Consumption Calculation (Section 1R21.b.1).

Licensee-Identified Violations

None

REPORT DETAILS

REACTOR SAFETY

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

1R21 Component Design Bases Inspection

The team selected risk-significant components and operator actions for review using information contained in the licensees probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater 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 reviewed design bases assumptions, calculations, and procedures. In some instances, the team performed calculations to independently verify the licensee's conclusions. The team also verified that the condition of the components was consistent with the design bases and that the tested capabilities met the required criteria.

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

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions because of modifications, and margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed performance test results; significant corrective actions; repeated maintenance; 10 CFR 50.65(a)1 status; operable, but degraded conditions; NRC resident inspector input of problem equipment; system health reports; industry operating experience; and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in-depth margins.

The inspection procedure requires a review of 15-20 risk-significant and low design margin components, 3-5 relatively high-risk operator actions, and 4-6 operating experience 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 check valve

• CV-1277 (CL), 'B' decay heat removal loop discharge to Make-Up Pump P-36C suction

• CV-1276 (CL), 'A' decay heat removal loop discharge to Make-Up Pump P-36A suction

• 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-50B inlet valve

• CV-3840 (CL), low pressure injection/Decay Heat Pump Bearing Cooler E-50A inlet

• 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 pressure pressure relief valves, low temperature over pressure motor-operated valve stop valves, 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 system pressure 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, dated June 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 (Unannounced Casualties) (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 Transformer Sizing Concerns

• Information Notice 2006-031: Inadequate Fault Interrupting Rating of Breakers

.1 Nonconservative Errors in Unit 2 Fuel Oil Consumption Calculation

Introduction:

The inspectors identified a noncited violation of 10 CFR Part 50 Appendix B, Criterion III, "Design Control", for Multiple Non-conservative Errors in a Unit 2 Emergency Diesel Generator Fuel Oil Consumption Calculation". Contributing factors to the errors were illegible reference data, inconsistently applied methodology, and inadequate calculation reviews.

Description:

Emergency diesel generator fuel oil consumption calculations should employ conservative assumptions to show that under worst-case conditions, there remains enough fuel oil storage capacity on site for the emergency diesel generators to fulfill their safety functions.

Arkansas Nuclear One, Unit 2, Calculation 91-E-0107-04, "Emergency Diesel Generator Fuel Oil Consumption," used the original diesel vendor factory test data as input to calculate the highest case consumption rate and the required fuel oil storage margin to meeting all of the safety analysis run time assumptions.

The safety analyses require run times based on 50, 100, and 110 percent diesel load ratings. The vendor Final Acceptance test that is used as input to the calculation ran the diesel at all three load ratings and recorded information at each one, including fuel oil consumption rate. The calculation methodology took the consumption rates recorded during the test and corrected them to worst-case scenarios to get a calculated worst-case consumption at each diesel load rating. These values where then used to calculate maximum run times at different ratings, given the various fuel oil volumes available in different safety analysis assumptions.

The vendor test data sheet was recorded by hand on the day of the test, October 26, 1979. The licensees official copy of this record is from an electronic system to which the original was either scanned or microfilmed. This test document carried a warning cover sheet indicating that parts of the document have been determined to be illegible and no better copy exists in the official records.

The main inputs to the calculation from the vendor data sheet are the recorded fuel consumption rates in lb/hr at each load rating. The vendor performed multiple runs of the diesel at each of the three load ratings. The calculation author attempted to pick the highest recorded consumption rate for each load rating as the calculation input, which was the most conservative approach. Upon close inspection of the document, it appears 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 the vendor. Looking at the unofficial copy, one of the 50 percent load rating runs had recorded 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 should have been used.

The calculation employed a correction factor to the consumption numbers taken from the test data in order to correct for worst-case fuel low heating values. The heating value recorded during the test was 19,678 BTU/hr. Arkansas Nuclear One test the fuel oil for low heating value. The calculation utilized the lowest low heating value the site had recorded in the past 5 years at the time of the calculation (1997) and found it to be 17,847 BTU/lb. The consumption values, at each load rating from the test data, were then multiplied by the ratio of these two heating values, which increased the vendor consumption rates by about 10 percent.

The inspectors reviewed the licensees fuel oil testing procedure and noted that although low heating values were recorded for information only, the procedure does place a limit on the range of acceptable values. If the sample result values fall outside this range, the procedure requires a condition report be written. According to the procedural limits, the lowest acceptable low heat value is 17,065 BTU/lb. The NRC inspectors challenged the use of 17,847 BTU/lb as the bounding value in the calculation, given that fuel with a low heating 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 in the original consumption numbers from the test data being increased by a 15 percent correction factor vice the original 10 percent.

This error reduces the calculated diesel run time margins for all scenarios in the calculation. The most sensitive, however, is the calculation of the 7-day safety analysis requirement for a single diesel train (one storage tank, one day tank) to operate at 50 percent rating after a design basis flood event. The original calculation resulted in only a 0.7 percent margin to meeting this 7-day requirement (7.05 days of run time available). Once the data sheet transcription error, which only affects the consumption rate 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 calculation when examining the consumption rate listed at 110 percent load. Instead of employing the same methods used at 50 and 100 percent load by taking the largest consumption rate at 110 percent load (1636 lb/hr) from the vendor data sheet, the licensee interpolated a smaller value of 1622 lb/hr from the data. Using the number from the data sheet reduces the margin in most of the calculation results, although no results were challenged in this case.

Also of concern, the licensee has not obtained actual data from recent diesel generator testing 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 the operating characteristics of the diesel, including fuel oil consumption rates.

When presented with these errors and the questioned ability to meet the safety analyses and design requirements, the licensee began scrutinizing the calculation. Eventually, the licensee found a vendor document from the test runs that called out the 19,678 BTU/lb heating value used in the calculation correction factor as a high heat value. High heating values account for the heat needed to vaporize the water in the fuel oil (heat of vaporization). This heat is not useful to the engine, so typically the heating value of interest is the low heat value, which does not include the heat of vaporization of water. However, the use of the high heat value in the calculation, while not technically correct, does add conservatism. In correcting that conservative error, the licensee used an appropriate low heat value, which reduced the correction factor enough to compensate for the nonconservative errors identified by the NRC. The licensee was able 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 a design bases emergency diesel generator fuel oil consumption calculation constituted a performance deficiency and a violation of 10 CFR Part 50, Appendix B, Criterion III. The violation was more than minor because it required the fuel oil volume calculations to be performed again to assure the accident analysis requirements were met. In accordance with Inspection Manual Chapter 0609, "Significance Determination Process,"

Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," the inspectors conducted a Phase 1 screening and determined the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss-of-operability in accordance with Part 9900, Technical

Guidance, "Operability Determination Process for Operability and Functional Assessment." The licensee entered this finding into the corrective action program as Condition Report ANO-2-2007-01325.

Enforcement:

Criterion III of Appendix B to 10 CFR Part 50 requires, in part, that measures shall be established for the identification and control of design interfaces and for coordination among participating design organizations. These measures shall include the establishment of procedures among participating design organizations for the review, approval, release, distribution, and revision of documents involving design interfaces. The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program.

Contrary to the above, as of September 17, 2007, the design control measures taken were not adequate with respect to Calculation 91-E-0107-04, Revision 2, which contained multiple errors that affected the emergency diesel generator fuel oil calculation results, some of which reduced the calculated margin to meeting design bases requirements. This violation is of very low safety significance and has been entered into the licensee's corrective action program as Condition Report ANO-2-2007-01325, and it is being treated as a noncited violation consistent with 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 Vortexing

Introduction:

The team reviewed the refueling water tank volume and level setpoint calculations, operating procedures, and the transfer to the reactor containment building sump, to determine if sufficient water exists in the tank to support the technical specification requirements. The refueling water tank was also reviewed to determine if sufficient water remains in the refueling water tank following transfer to the containment sump during a loss-of-coolant accident to prevent drawing air into the safety injection and containment spray pumps. The time for automatic transfer was reviewed by the team to determine if adequate water remained in the take to allow for swap-over to the containment sump.

The team also reviewed the quality condensate storage tank volume requirement and the tank level during the transfer from the quality condensate storage tank to the service water system by the emergency feedwater system. The team reviewed the calculations for Units 1 and 2 and operating procedures for quality condensate storage tank transfer to determine if adequate water exists during the transfer operation. The team reviewed the calculation of required water level in the tank to facilitate the 30 minute transfer from the quality condensate storage tank to the service water system.

Description:

The automatic refueling water tank recirculation actuation system setpoint is documented in Calculation 93-EQ-2001-03. This includes the instrument uncertainty considerations, as well as development of the recirculation actuation system setpoint, primarily to show how safety analysis volumes are achieved. For the refueling water tank, the team found that although sufficient water exists to support the technical specification requirements, Arkansas Nuclear One relies on a vortex suppressor to assure that there will be no air drawn into the safety injection and containment spray pumps prior to swap over. The refueling water tank level that swap over occurs was established based on the use of a vortex suppressor. However, the vortex suppressor does not have either analytical or test data to support its use. A corrective action document was issued by Arkansas Nuclear One to evaluate the design and determine if adequate water level exists. Arkansas Nuclear One prepared an operability evaluation and determined that the plant remains operable in the current condition.

The quality condensate storage tank level setpoint was established to provide a 30 minute supply of water to facilitate the emergency operating procedure mandated manual 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 the water level during the transfer to the service water system is adequate. The swap over level was established based on the use of a vortex suppressor. Although the setpoint level is supported by Calculations ANO-1: CALC-90-E-0116-07 Setpoint T.6 and ANO-2: CALC-90-E-0116-01 Setpoint T.18, the vortex suppressor does not have either analytical or test data to support its use. Arkansas Nuclear One issued a corrective action to analyze both vortex suppressors. The NRC will evaluate the adequacy of the refueling water tank and quality condensate storage tank transfer setpoints when the licensee has completed testing and analysis of the vortex suppressers. (URI 05000313; 368/2007008-02; ).

Analysis:

The NRC will complete a significance determination, if warranted, when closing out the unresolved item.

Enforcement:

The NRC will consider enforcement, if necessary, when closing out the unresolved item.

.3 Water Storage Tanks

a. Inspection Scope

The team evaluated the instrumentation associated with various water storage tanks to ensure there was adequate capability (volume, chemical concentration and temperature)to properly fulfill the design provisions stated in the Updated Final Safety Analysis Report and other design commitments for each of the evaluated tanks. The tanks that were reviewed included the Unit 1 borated water storage tank, the Unit 2 refueling water tank and the common qualified condensate water storage tank. The team reviewed the calculations for the instrumentation utilized for level and temperature monitoring related to these storage tanks in order to verify that uncertainties and scaling properties had been properly incorporated into the indication and control circuitry devices. The team also 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 water storage tank, the team noted that the last calibration of the temperature instrumentation had been performed on February 12, 2003. Because this temperature instrumentation is used to verify Unit 1 Technical Specification Surveillance Requirement 3.5.4.1, the team questioned the appropriateness of the long interval since its last calibration. The licensee investigated the calibration history for the borated water storage tank temperature Transmitter TT-1413, and found that the instrument had been removed from the routine, 3-year interval, testing and calibration program on July 20, 2006. The licensee initiated Condition Report ANO-1-2007-02041 to evaluate the cause of the instrument being removed from the routine testing and calibration program and to implement corrective actions.

The team verified that the borated water storage tank level instrumentation and that the comparable Unit 2 refueling water tank level and temperature instrumentation continued to be routinely tested and calibrated. A calibration check of the borated water storage tank temperature transmitter and instrumentation string was completed in accordance with Repetitive Maintenance Task No. 10638 on September 19, 2007. The instrumentation was found to be within acceptable limits.

b. Findings

No significant findings were identified.

.5 Station Batteries

a. Inspection Scope

The team evaluated Units 1 and 2 safety-related station batteries to ensure there was adequate capability to fulfill the design provisions stated in the Updated Final Safety Analysis Reports and other design commitments. The team reviewed the dc voltage requirement calculations for both units to verify that sufficient voltage would be available at the terminals of selected loads to ensure their proper operation.

The team reviewed the testing procedures and the results of recently completed battery capacity tests to verify that the batteries could perform the safety functions described in the Updated Final Safety Analysis Report. The team questioned several portions of the procedures that were implemented on both Units 1 and 2 to verify that the methodology was in accordance with manufacturer recommendations and IEEE-450 guidance. The team also reviewed selected condition reports that had been initiated for problems identified during testing of the station batteries to verify that appropriate actions had been implemented.

The team reviewed the licensees studies and calculations pertaining to the ability of each unit to cope with a station blackout. The team verified that the licensees calculations concluded that adequate voltage would be available to perform such functions as 4160V circuit breaker actuations at the end of the station blackout coping period. As part of the station blackout reviews, the team reviewed the offsite power supply system to ensure that an instability on one of the systems (500 or 161 kV) would not result in the loss of the other system. The team also inspected the battery systems for both portions of the switchyard to verify that proper maintenance was being conducted.

b. Findings

No significant findings were identified.

.6 Emergency Diesel Generator Field Flashing

a. Inspection Scope

During a walkdown of the electrical components installed in the facility, the team questioned the function of the small 24V batteries installed in the Unit 2 emergency diesel generator rooms. The team was informed that the batteries (two 12V batteries connected in series) had been installed to provide an emergency source of power for field flashing. The team reviewed the information the licensee had received from the Unit 2 emergency diesel generator vendor stating that 12V would be adequate for field flashing of the Unit 2 emergency diesel generator s.

The team also questioned the field flash requirements for the Unit 1 emergency diesel generator. The team was informed that a test had been conducted that verified residual magnetism was adequate for field flashing of the Unit 1 emergency diesel generators and, therefore, no separate source of power was required. The team reviewed the temporary test procedure and noted that generator voltage levels were achieved without applying a source of field flashing current.

b. Findings

No findings of significance were identified.

.7 Motor-Operated Valve

a. Inspection Scope

The team selected motor-operated valves installed in both units to determine if the valves and their actuators could properly fulfill the design functions discussed in the Updated Final Safety Analysis Report and other design documents. The team selected the containment sump outlet valves from both units, the Unit 1 borated water storage tank 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 degraded voltage analysis and related calculations to ensure that degraded voltage levels had been utilized in the determination of actuator motor torque capabilities.

The team reviewed the schematic diagrams for the selected motor-operated valves to verify that the required actuation and interlock signals had been appropriately incorporated in the circuitry. The team also evaluated the circuitry related to the motor thermal overload protective feature to verify that the feature was in accordance with the licensees commitments to Regulatory Guide 1.106, "Thermal Overload Protection for Electric Motors on Motor-Operated Valves." Regulatory Position C.1 of Regulatory Guide 1.106 states that the thermal overload feature should be bypassed during accident conditions; Regulatory Position C.2 of the Regulatory Guide states that the trip setpoint of the thermal overload protection devices should be established with all uncertainties resolved in favor of completing the safety-related action. The team determined that Arkansas Nuclear One, Unit 1, was committed to Regulatory Position C.2 of Regulatory Guide 1.106 and that Unit 2 was committed to Regulatory Position C.1 of the Regulatory Guide.

The team verified that the schematic diagrams for the selected Unit 2 motor-operated valves indicated that the thermal overload protection feature was bypassed under accident conditions. The team reviewed the licensees standards for sizing motor-operated valve thermal overload heaters and noted that the standard referenced the provisions of Regulatory Position C.2 of Regulatory Guide 1.106 as one of the considerations in the motor-operated valve heater selection process. The team evaluated the thermal overload heater size installed in three of the selected Unit 1 motor-operated valves and verified that the installed heater actuation setpoints were higher than the motor amperage recorded during valve testing.

b. Findings

No findings of significance were identified.

.8 Borated Water Storage Tank Volume (Unit 1)

a. Inspection Scope

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

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 discharge valves to evaluate if the valves will open in accordance within the time requirements and close during the transfer from the tanks to the containment sump. The team performed the following activities: reviewed corrective actions, interviewed the system engineers for Units 1 and 2 to determine if the valves had operating or maintenance issues, and interviewed the motor-operated valve engineer to determine if the valves had adequate design margin. The valve calculations, corrective actions and inservice test results were reviewed.

b. Findings

No findings of significance were identified.

.10 Unit 1 and 2 Containment Sump Discharge Valves

a. Inspection Scope

The team reviewed the containment sump inboard and outboard discharge valves to assure that the valves would function during a emergency core cooling system event.

The team reviewed past corrective actions, valve design calculations and test results and interviewed the systems engineers for Units 1 and 2 to determine if the valves had operating or maintenance issues. Particular focus was provided by the team for corrective actions for closing of the Unit 2 valves and bypass leakage. The team reviewed the assumptions for the control room and off-site dose analysis that was used in the corrective action.

b. Findings

There were no findings of significance identified

.11 High Pressure Safety Injection Pumps (Unit 2)

a. Inspection Scope

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

b. Findings

There were no findings of significance identified.

.12 Low Pressure Decay Heat Removal System Piping (Unit 1)

a. Inspection Scope

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

b. Findings

There were no findings of significance identified

4OA6 Meetings, Including Exit

On September 21, 2007, the team leader presented the preliminary inspection results to Mr. Tim Mitchell, Site Vice-President, and other members of the licensees staff. The licensee acknowledged the findings during each meeting. While some proprietary information was reviewed during this inspection, no proprietary information was included in this report.

s:

1. Supplemental Information 2. Initial Information Request

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

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

Opened

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

Attachment 1