IR 05000327/2007006

IR 05000327-07-006 and 05000328-07-006, on June 4-July 13, 2007, Sequoyah Nuclear Plant - Component Design Bases Inspection

ML072890690
Person / Time
Site:	Sequoyah
Issue date:	10/11/2007
From:	Binoy Desai NRC/RGN-II/DRS/EB1
To:	Campbell W Tennessee Valley Authority
References
IR-07-006
Download: ML072890690 (36)

Text

October 11, 2007

SUBJECT:

SEQUOYAH NUCLEAR PLANT - COMPONENT DESIGN BASES INSPECTION

- NRC INSPECTION REPORT 05000327/2007006 AND 05000328/2007006

Dear Mr. Campbell:

On July 13, 2007, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Sequoyah Nuclear Plant Units 1 and 2. The enclosed inspection report documents the inspection findings which were discussed during the preliminary exit meeting on July 13, 2007,

and the final exit meeting on September 10, 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 Commissions rules and regulations and with the conditions of your license.

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

Based on the results of this inspection, the inspectors identified four findings of very low safety significance (Green). These findings were determined to involve violations of NRC requirements. However, because of the very low safety significance and because each was entered into your corrective action program, the NRC is treating the findings as non-cited violations consistent with Section VI.A.1 of the NRCs Enforcement Policy. If you deny these non-cited violations you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the United States Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Sequoyah Nuclear Plant.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS).

TVA

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Binoy Desai, Branch Chief Engineering Branch 1 Division of Reactor Safety Docket Nos.: 50-327, 50-328 License Nos.: DPR-77, DPR-79

Enclosure:

NRC Inspection Report 05000327/2007006 AND 05000328/2007006 w/Attachment: Supplemental Information

REGION II==

Docket Nos.:

50-327, 50-328 License Nos.:

DPR-77, DPR-79 Report Nos.:

05000327/2007006, 05000328/2007006 Licensee:

Tennessee Valley Authority (TVA)

Facility:

Sequoyah Nuclear Plant Location:

Sequoyah Access Road Soddy-Daisy, TN 37379 Dates:

June 4 - July 13, 2007 Inspectors:

R. Taylor, Lead Inspector R. Moore, Senior Reactor Inspector W. Fowler, Reactor Inspector J. Rivera-Ortiz, Reactor Inspector S. Kobylarz, Contractor T. Tinkel, Contractor M. Lewis, Inspector Trainee Approved by:

Binoy Desai, Branch Chief, Engineering Branch 1 Division of Reactor Safety

Enclosure

SUMMARY OF FINDINGS

IR 05000327/2007006; 05000328/2007006; 06/4/2007 - 06/8/2007, 06/18/2007 - 06/22/2007, 07/9/2007 - 07/13/2007; Sequoyah Nuclear Plant, Units 1 and 2; Component Design Bases Inspection.

This inspection was conducted by a team of four NRC inspectors and two NRC contractors.

Four green findings, all of which were non cited violations, were identified during this inspection.

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

!

Green.

The team identified a violation of Technical Specification 6.8.1 associated with Tennessee Valley Authoritys (TVA) failure to develop a procedure that would provide periodic inspection and replacement of the emergency diesel generator room electrical panel ventilation air filters. This failure resulted in a fourteen year period between 1993 and present, for which the filters were not inspected and replaced for all four diesel generator rooms.

The finding is greater than minor because it is associated with the Mitigating Systems cornerstone attribute procedure quality and affects the cornerstones objective of ensuring the availability, reliability, and operability of the emergency diesel generators to perform their safety function during an event, such as a loss of offsite power (LOSP). In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the team conducted a Phase 1 SDP screening and determined the finding was of very low safety significance (Green). This finding was entered into Sequoyahs Corrective Action Program (CAP) under Problem Evaluation Report (PER) 125944 and actions have been taken to replace the existing air filters and procedural changes have been made to ensure filter inspection and replacement will be performed on a periodic basis. (Section 1R21.2)

Cornerstone: Mitigating Systems

!

Green.

The team identified a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, associated with TVAs use of non-conservative design input values in design calculations. TVAs failure to use appropriate inputs in design calculations resulted in a significant increase in the calculated maximum room temperatures for the Emergency Core Cooling System (ECCS) pump rooms as well as the Turbine Driven Auxiliary Feedwater (TDAFW) pump room. This increase in the calculated maximum room temperatures had the potential for affecting the operability of safety related components in the ECCS and TDAFW pump rooms.

This finding is more than minor because if left uncorrected the use of incorrect design input values could become a more serious safety concern as many other safety-related design calculations rely upon these design outputs. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the team conducted a Phase 1 SDP screening and determined the finding was of very low safety significance (Green). This finding was entered into Sequoyahs CAP under PER 127414. (Section 1R21.2)

Cornerstone: Initiating Events

!

Green.

The team identified a violation of TS 6.8.1 for failure to establish an adequate procedure for reactor coolant system reduced inventory/mid-loop operations.

Specifically, procedure 0-GO-13, Reactor Coolant System (RCS) Drain and Fill Operations, Rev. 57, was not adequate in that it did not establish adequate actions to maintain continuous RCS level indication during all possible plant conditions while in the reduced inventory/mid-loop configuration, specifically Loss of Offsite Power (LOSP).

Additionally, the procedure did not establish contingency actions to recover power to the Mansell level indication systems or provide guidance for alignment of an alternate RCS level indication mechanism within the 30 minutes for which power was still available from the Mansell system unit battery.

This finding is more than minor because it impacts the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown. In accordance with NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations the team conducted a Phase 3 SDP screening and determined the finding was of very low safety significance (Green). This finding was entered into the Sequoyah corrective action program as PER 125906.

(Section 1R21.3)

Cornerstone: Initiating Events

!

Green.

The team identified a violation of TS 6.8.1 for failure to establish an adequate abnormal operating procedure for the Residual Heat Removal System (RHR) system malfunctions during shutdown conditions. Procedure AOP-R.03, RHR System Malfunction, Rev. 17, was not adequate in that it did not establish adequate actions to restore RHR cooling following isolation of an RHR leak during hot shutdown (Mode 4)operations. The instruction provided in the procedure could result in a total loss of RHR cooling capability during Mode 4 conditions if an RHR leak occurred.

This finding is more than minor because it impacts the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown in that the loss of RHR pumps would increase the likelihood of a loss of RHR cooling. In accordance with NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations the team conducted a Phase 2 SDP screening and determined the finding was of very low safety significance (Green). This finding was entered into Sequoyahs CAP under PER 125844. (Section 1R21.3)

B.

Licensee-identified Violations None

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Mitigating Systems and Barrier Integrity

1R21 Component Design Bases Inspection

.1 Inspection Sample Selection Process

The team selected risk significant components and operator actions for review using information contained in the licensees Probabilistic Risk Assessment (PRA). In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum value greater than 1E-6. The components selected were located within several safety related systems. The sample selection included 18 components, 5 operator actions, and 5 operating experience items.

Additionally, the team reviewed two modifications by performing activities identified in IP 71111.17, Permanent Plant Modifications, Section 02.02.a. and IP 71111.02, Evaluations of Changes, Tests, or Experiments.

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modification, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed performance test results, significant corrective action, repeated maintenance, maintenance rule (a)1 status, Regulatory Issue Summary 05-020 (formerly GL 91-18)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. An overall summary of the reviews performed and the specific inspection findings identified are included in the following sections of the report.

.2.1 Residual Heat Removal (RHR) Pumps

a. Inspection Scope

The team reviewed the Updated Final Safety Analysis Report (UFSAR), design basis documents, Emergency Operating Procedures (EOP), and calculations to verify that the Residual Heat Removal (RHR) pumps have adequate Net Positive Suction Head (NPSH) in order to perform their design basis function of providing low head safety injection flow during containment recirculation. The team reviewed calculations, which are the basis for providing the operators guidance in the EOPs that define 100 minutes as the maximum allowed time to operate the RHR pumps on minimum flow without Component Cooling System (CCS) flow to the RHR heat exchanger. Situations where this would occur would be during events, such as, a small break loss of coolant accident (SBLOCA). Inputs into the calculations were reviewed and compared against the current plant configuration to verify any changes to the facility have not adversely impacted the calculations results. Equipment used to monitor the performance capabilities of the RHR pump mechanical seal cooler were reviewed to verify the capability to promptly identify degrading cooler performance. Walkdowns were performed to verify component degradation is being prevented and that the physical condition is being maintained.

b. Findings

No findings of significance were identified.

.2.2 Turbine Driven Auxiliary Feedwater (TDAFW) Pumps

a. Inspection Scope

The team reviewed design criteria, UFSAR calculations, mechanical flow and control diagrams, electrical and logic drawings, and system operating procedures, to verify that the AFW system design basis have been adequately implemented and maintained. The team reviewed pump curves, system surveillance testing, and comprehensive testing records to verify that the installed TDAFW pumps are capable of developing the required flow and pressure for accident conditions.

The team reviewed environmental qualification information for the equipment inside the TDAFW pump rooms to verify that the equipment has been demonstrated to operate during the expected environmental conditions of a postulated accident and a station blackout event. The team also reviewed design and testing documentation for the TDAFW pump room ventilation to verify that adequate room cooling is available to maintain design temperatures.

b. Findings

Introduction.

The team identified a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, associated with TVAs failure to use appropriate assumptions in design calculations at the Sequoyah Nuclear Plant. TVAs failure to use appropriate assumptions in design calculations resulted in a significant increase in the calculated maximum room temperatures for the Emergency Core Cooling System pump rooms as well as the TDAFW pump room. This increase in the calculated maximum room temperatures led to a reasonable doubt about the operability of components in the affected rooms.

Description.

Design Criteria Document SQN-DC-V-13.9.3, Auxiliary Building Ventilation and Cooling, Table T1.34 established the design temperatures for the TDAFW pump room for normal and abnormal conditions. This document established a maximum room temperature of 110° F for an abnormal operational event. TVA calculation SQN-31C-D053-EPM-RG-060987, Revision 2 was developed, in part, to determine if the capacity of the installed direct current (DC) fan equipment for the TDAFW pump was adequate to maintain the maximum design temperature of 110° F specified in the design criteria document. The calculation determined the air temperature needed to maintain the abnormal design temperature based on the heat input load calculated in SQN-31C-D053-EPM-DLM01-030887, HVAC Cooling Load Calculation: Aux. Bldg. TDAFW pump, Revision 2 and determined an inlet air temperature of 80° F. Based on a preliminary calculation, the installed fan would maintain the TDAFW pump room at 131° F instead of 110° F. This issue was entered into the CAP as PER 126928. A review of the environmental qualification (EQ) list of equipment inside that room indicated that the equipment is qualified to at an ambient temperature of 215° F which is higher than the 110F° specified in the design criteria document.

Another design calculation with inappropriate assumptions was Engineered Safety Features (EFS) room cooler calculation, 30-DO53-EPM-BVC-052788, Emergency Raw Cooling Water (ERCW) River Water Temperatures Effect on ESF Coolers, Rev. 7. The incorrect assumption was the use of input cooler air flow rates that were higher than the minimum design flow rate. In this case, the non conservative nominal value of 4933 cubic feet per minute (CFM) was used as opposed to the minimum design airflow rate of 4439 CFM in order to calculate maximum accident room temperature. Deficient assumptions included using an ERCW supply temperature and flow rate that were not worst case for the calculations. As a result of the use of non conservative input values in the ESF maximum room temperature calculations, the installed room coolers were not capable of maintaining the analyzed temperatures for the ESF rooms. The licensee performed a preliminary calculation which showed a significant reduction in margin between the calculated maximum room temperatures and design limits. In the case of the 1B-B Centrifugal Charging Pump (CCP) room, the design temperature limit was exceeded by

.6 degrees.

The team determined the 1B-B CCP to be operable because of additional margin available with respect to environmental qualification.

Analysis.

The TVA's failure to use appropriate assumptions for ESF and TDAFW pump maximum room temperature calculations is a performance deficiency associated with the Mitigating Systems cornerstone. This finding is more than minor because if left uncorrected problems in Design Control could lead to a more serious safety concern as many other safety-related design calculations rely upon these design outputs. This finding was reviewed for cross-cutting aspects and none were identified.

Enforcement.

10 CFR 50, Appendix B, Criterion III, Design Control specifies, design control measures shall provide for verifying the adequacy of design by the use of alternate or simplified calculational methods. Contrary to the above, TVA calculations did not provide adequate measures for verifying the adequacy of the design of the ESF room coolers or TDAFW ventilation system. This finding was entered into Sequoyah's Corrective Action Program under PER 127414. Because this finding is of very low safety significance and was entered into Sequoyah Nuclear Plants (SNP) corrective action program as PER 127414, this violation is being treated as a non-cited violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000327,05000328/2007006-01, Failure To Use Appropriate Assumptions in Design Calculations)

.2.3 RHR Minimum Flow Valve, 1-FCV-74-12

a. Inspection Scope

The team reviewed the UFSAR, system descriptions, and evaluations associated with the required setpoint for positioning the RHR minimum flow valve in the full open and closed positions to verify the RHR pumps will not dead head and that full RHR flow will be achieved at the closure setpoint. RHR flow tests and minimum flow valve actuation tests were reviewed to verify the necessary flow rates are being achieved and that the minimum flow valve was responding at the required points and achieving its desired position. The team reviewed the operating parameters recorded from previous RHR surveillance testing to assess the potential for development of air voids that could degrade system performance.

System walkdowns were performed to verify the valve and its supporting electrical connections have not been impacted by degradation and that its pressure boundary integrity is being maintained. Motor-operated valve actuator test (MOVAT) testing was observed during the inspection to verify the level of technical knowledge was appropriate and procedures are being adhered to.

b. Findings

No findings of significance were identified.

.2.4 1-FCV-63-8 (RHR Heat Exchanger A Discharge to CCP Suction) and 1-FCV-63-11

(RHR Heat Exchanger B Discharge to SI Pump Suction)

a. Inspection Scope

The team reviewed design criteria documents, UFSAR, calculations, mechanical flow and control diagrams, electrical and logic drawings, and system operating procedures, to verify that the ECCS design basis for recirculation mode have been adequately implemented and maintained. The team reviewed calculations developed to determine the thrust and torque required to open and close the valves under accident conditions along with the evaluation of the installed actuator. Maintenance, modification, and corrective action history of these MOVs was reviewed to assess the reliability of these components and the licensee capability to identify degradation.

The team also reviewed environmental qualification documentation of these MOVs to verify they have been demonstrated to operate during the expected ambient conditions of a postulated accident. Additionally, the team reviewed Thermal Overload (TOL)information of these MOVs in order to verify that TOL protection will not adversely impact the valve design function. The teams TOL review included an independent calculation to assess the adequacy of the TOL heaters selected for these MOVs. The team reviewed logic and electrical diagrams to verify that permissive interlocks and valve circuitry met design basis requirements. The team walked down the valves to assess material condition and correct position.

b. Findings

No findings of significance were identified.

.2.5 1-FCV-63-72 (RHR 1A-A Pump Suction from Containment Sump) and 1-FCV-63-73

(RHR 1B-B Pump Suction from Containment Sump )

a. Inspection Scope

The team reviewed design criteria documents, UFSAR, calculations, mechanical flow and control diagrams, electrical and logic drawings, and system operating procedures to verify that the ECCS design basis for recirculation mode have been adequately implemented and maintained. Maintenance, modification, and corrective action history was reviewed to assess the reliability of these components and the licensee capability to identify degradation. The team reviewed calculations developed to determine the thrust and torque required to open and close the valves under normal, testing, and accident conditions along with the evaluation of the installed actuator capability to deliver the required thrust and torque to verify that these MOVs could function under accident conditions.

The team also reviewed environmental qualification documentation to verify they have been demonstrated to operate during the expected ambient environmental conditions of a postulated accident. Additionally, the team reviewed Thermal Overload (TOL)information of these MOVs in order to verify that TOL protection will not adversely impact the valve design function. The teams TOL review included an independent calculation to assess the adequacy of the TOL heaters selected. The team reviewed logic and electrical diagrams to verify that permissive interlocks and valve circuitry satisfied the design criteria documentation. The team also reviewed stroke test results to assess valve performance and interlock testing. The team conducted walkdowns to assess material condition and correct position.

b. Findings

No findings of significance were identified.

.2.6 Refueling Water Storage Tank(RWST) Supply Valves(FCV-63-5, FCV-63-47 and FCV-

63-48) to Safety Injection(SI) Pumps

a. Inspection Scope

The team reviewed design basis information including functional requirements identified in the UFSAR, System Design Criteria Documents (SDCD), and selected drawings:

FCV-63-5, FCV-63-47 and FCV-63-48. Valve stem thrust and MOV calculations, setup sheets, and test results were reviewed to verify proper actuator sizing and torque switch bypass settings. Thermal overloads settings were reviewed to determine whether they were installed and adequately sized. Calculations and surveillance test results were reviewed to identify whether these gate valves were susceptible to pressure locking and whether they could operate under worst case design basis and accident system pressure and differential pressure.

b. Findings

No findings of significance were identified.

.2.7 Refuling Water Storage Tank (RWST) to RHR Suction Valve, 1-FCV-74-21

a. Inspection Scope

The team reviewed the UFSAR, system descriptions, and in-service testing requirements to identify the design basis functions of the RWST suction isolation valve.

Control circuitry documents were reviewed to verify valve operation and interlocks with the containment sump valves are as described by the licensing basis documents.

Calculations associated with required thrust to open or close the valve were verified to not exceed the motor's output capabilities. Documentation for the periodic stroke test was reviewed to verify the required stoke time is being met as required by design documents associated with Generic Letter 89-10. Valves associated with preventing a potential path from the containment sump to the RWST were reviewed to verify they will be in the closed position or procedurally positioned closed in the event the suction isolation valve fails to fully seat and that periodic testing has validated this capability.

Walkdowns were performed to verify correct valve position and component degradation.

Visual inspections of the RHR pump sumps were performed to verify any leakage would be appropriately detected by sump level indication and that the inventory would be removed.

b. Findings

No findings of significance were identified.

.2.8 Essential Raw Cooling Water (ERCW) Pumps

a. Inspection Scope

The team reviewed design basis information including functional requirements identified in the UFSAR, System Design Criteria Documents (SDCD), and selected drawings to identify design basis requirements. Inservice test procedure acceptance limits and the system design curves were reviewed and compared to assure minimum acceptance criteria were satisfied. The ERCW MULTIFLOW hydraulic model and pump test results were reviewed to verify the pumps were capable of supplying required head and flow under worst case design basis and accident conditions. Test results and material history were reviewed to evaluate the performance history of the pumps and to identify failures, off-nominal performance, and any trends of degrading performance. Problem Evaluation Reports (PERs) and flow verification test data were reviewed to identify any system flow balance issues and to compare flow test data with calculated results. Pump motor electric power requirements were examined and compared with load values used in the diesel generator load analysis. A system walkdown was performed in the pump house area to observe the material condition of the pump motors, strainers, and associated piping.

b. Findings

No findings of significance were identified.

.2.9 ERCW Strainers

a. Inspection Scope

The team reviewed design basis information including functional requirements identified in the UFSAR, SDCD, vendor documents, and selected drawings to identify design basis requirements including the flow resistance characteristics of the strainers. ERCW system calculations were reviewed to determine whether the strainer flow resistance characteristics were modeled appropriately and that worst case river levels were accounted for in determining minimum flow requirements for design basis and accident conditions. The team reviewed licensee programs and procedures used to monitor and control fouling of ERCW components from organics such as clams. Circulating water (CW) system PERS were reviewed to determine whether any issues might be generic to the ERCW system. A system walkdown was performed in the pump house area to observe the material condition of the pump motors, strainers, and associated piping.

b. Findings

No findings of significance were identified.

.2.10 ERCW Room Cooler for Safety Injection Pump (SIP)-1A

a. Inspection Scope

The team reviewed design basis information including functional requirements identified in the UFSAR, SDCD, and selected drawings to identify design basis requirements for the SIP 1A room cooler. Vendor drawings were reviewed to identify whether the heat transfer rating of the cooler was appropriate to transfer the heat load found in the SIP 1A room heat load calculation. The team inspected procedures used to periodically measure and assure cooler air and water flows rates were adequate to transfer the required heat load of the room. Trend reports maintained by the System Engineer were reviewed to confirm that cooler performance was being monitored and that corrective action was taken to address adverse trends. PERs and material history records were reviewed to identify any instances of tube leakage and to verify the associated corrective actions were effective.

b. Findings

No findings of significance were identified.

.2.11 Charging Pump Isolation from RWST, 1-LCV-062-136

a. Inspection Scope

The team reviewed design criteria documents, UFSAR, calculations, mechanical flow and control diagrams, electrical and logic drawings, and system operating procedures; to verify that the ECCS design basis for safety injection mode have been adequately implemented and maintained. Maintenance documentation was reviewed to determine that MOVs were periodically tested and inspected to ensure that the design function was maintained.

The team also reviewed environmental qualification documentation to verify it has been demonstrated to operate during the expected environmental conditions of a postulated accident. Additionally, the team reviewed TOL information in order to verify that TOL protection will not adversely impact the valve design function. The team reviewed logic and electrical diagrams to verify that permissive interlocks and valve circuitry satisfied the design criteria documentation. The team also reviewed stroke test results to assess valve performance and interlock testing. The team walked down the valve to assess material condition and verify it is in its required position.

b. Findings

No findings of significance were identified.

.2.12 ERCW to TDAFW Supply Valve, 1-FCV-03-136B

a. Inspection Scope

The team reviewed design basis information, such as, functional requirements identified in the Updated Final Safety Analysis Reports (UFSAR), System Design Criteria Documents (SDCD), and selected drawings for this motor operated valve (MOV), FCV-03-136B to identify design basis requirements. Valve stem thrust and MOV calculations, setup sheets, and test results were reviewed to verify proper actuator sizing and torque switch bypass settings. Thermal overloads settings were reviewed to determine whether overloads were installed and adequately sized. Surveillance instructions were reviewed to identify whether they contained requirements for testing interlocks and automatic actuation of the valve. Calculations and surveillance test results were reviewed to identify whether this gate valve was susceptible to pressure locking and whether there was reasonable assurance it could operate under worst case design basis and accident system pressure and differential pressure. With respect to the ERCW piping upstream of this valve, the team reviewed the program and procedures used to prevent organics such as clams and other debris from accumulating and entering the suction side of the TDAFW pump when aligned to the ERCW system.

AFW system hydraulic calculations were also reviewed to determine whether adequate net positive suction head was available when the TDAFW pump is aligned to the ERCW supply header via this valve. System health reports, PERS, and material history were reviewed to identify failures or off-nominal operation and to verify any associated corrective actions. A system walkdown was performed to observe the external material condition of the valve.

b. Findings

No findings of significance were identified.

.2.13 Emergency Diesel Generator (EDG) Ventilation Fans and Dampers

a. Inspection Scope

The team reviewed the UFSAR, design criteria documents, operator lesson plans, fire protection plan, and corrective action program documents to verify the EDG building ventilation system can perform its design basis function of providing adequate cooling during diesel engine operation, adverse weather conditions and ability to perform its CO2 suppression system functions. Procedural changes associated to ensure the ventilation dampers will remain open during periods of cold weather to prevent damage due to environmental pressure changes associated with tornadic activities were reviewed to verify they did not adversely affect other system functions, such as fire protection.

Electrical diagrams associated with the diesel building exhaust fan were reviewed to verify their capability to perform their design function of remaining off during a CO2 initiation signal. This ensure adequate concentration and hol time even if a diesel start signal comes in. Backup fan controls were reviewed to verify backup start signals would be received even if the primary fan trips due to thermal overload or a downstream electrical fault occurs upon a EDG start signal.

Walkdowns were performed to verify ventilation components have not degraded over time due to debris formation in ductwork. Ductwork connections were inspected to verify bypass flows that could inhibit proper ventilation paths were not present. The team reviewed preventative maintenance for the EDG room electrical panel ventilation components which included the air filter and cooling fan.

b. Findings

Introduction.

The team identified a violation of Technical Specification 6.8.1 associated with TVA's failure to develop a procedure that will provide periodic inspection and replacement of the EDG room electrical panel ventilation air filters at the Sequoyah Nuclear Plant. This resulted in a fourteen year period between 1993 and present, for which the filters were not inspected and replaced for all four diesel generator rooms.

Description.

The diesel generator room ventilation system consists of one air supply fan, one air duct that splits to provide ventilation air to the generator and the electrical panels, and one air filter located in the ductwork before the electrical panels. The purpose of the filter is to provide a clean supply of air in order to prevent debris accumulation in the panel, which could affect various electrical components. In the event the filter becomes clogged, due to long-term use, internal temperature limits for various electrical components associated with generator excitation controls could be impacted. This would affect the long-term reliability of the EDG to provide adequate voltage and power to its associated bus during its mission time.

Analysis.

TVA's failure to develop a procedure that will provide periodic inspection and replacement of the EDG room electrical panel ventilation air filter is a performance deficiency associated with the Mitigating Systems cornerstone. The finding is greater than minor because it is associated with the Mitigating Systems cornerstone attribute of procedure quality and affects the cornerstones objective of ensuring the availability, reliability, and operability of the emergency diesel generators to perform their safety function during an initiating event, such as, a loss of offsite power. This finding was reviewed for cross-cutting aspects and none were identified.

Enforcement.

Technical Specification 6.8.1, "Procedures & Programs", requires in part that written procedures be established, implemented and maintained per Regulatory Guide (RG) 1.33, Rev. 2. Appendix A of RG 1.33 states that procedures for performing maintenance shall be covered by written procedures. Contrary to the above, TVA did not develop procedures for replacing air filters for the EDG room electrical panel.

Because this finding is of very low safety significance and was entered into SNPs corrective action program as PER 125944, this violation is being treated as a non-cited violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000327,05000328/2007006-02, No Procedure for Inspection/Replacement of DG Local Electrical Panel Filters)

.2.14 480V Diesel Aux Boards (2B1-B)

a. Inspection Scope

The team reviewed the design basis documentation for the 480V Diesel Aux Boards to identify the analyzed sizing for worst case load and short circuit conditions and to verify the design basis requirements were consistent with the equipment specifications and vendor information. Maintenance and test activity was reviewed to verify that equipment degradation was monitored and to review equipment availability. A field walkdown of the diesel auxiliary boards was conducted to observe the general material condition and to verify that the board arrangement was in accordance with the single line diagram.

The coordination calculation was reviewed to verify breaker setpoint determination in is accordance with design basis conditions. Selected loads were reviewed to verify that the thermal overload relay heaters selected by the coordination calculation were sized in accordance with design requirements.

b. Findings

No findings of significance were identified.

.2.15 480V Shutdown Boards (1A2-A)

a. Inspection Scope

The team reviewed the design basis documentation for the 480V Shutdown Boards to identify the analyzed sizing for worst case load and short circuit conditions and to verify the design basis requirements were consistent with the equipment specifications and vendor information. Maintenance and test activities were reviewed to verify that equipment degradation was monitored and to review equipment availability. A field walkdown of the shutdown boards was conducted to observe general material conditions. The relay coordination calculation was reviewed to verify setpoint determination and incorporation of setpoint information into relay calibration procedures.

Relay calibration test results were reviewed to verify that any out of tolerance conditions were identified and resolved. Relay settings were verified in the field for consistency with the design calculations and the calibration test procedures.

b. Findings

No findings of significance were identified.

.2.16 Emergency Diesel Generator (1A-A) Starting Circuit

a. Inspection Scope

The team reviewed the EDG control circuit schematic diagrams to identify critical start circuit components. The team reviewed corrective maintenance history on critical start components for repeat problems and unavailability. A walkdown of the EDG control components was performed to observe material condition. The team reviewed the EDG modification history to identify changes and their effect on the design basis for the equipment.

The team reviewed the design analyses for the capability of the EDGs 125 VDC battery for design basis conditions. The battery capacity calculation was reviewed to identify the design requirements for the battery and support equipment. Battery testing and inspections were reviewed to assess the licensees actions to verify and maintain the design capability of the battery. The maintenance history on the battery and support equipment was reviewed to verify battery availability. A field verification was performed to observe the material condition of the battery and verify the battery load analysis designated loads were correctly based on the arrangement of the distribution panels.

Additionally, alarms and indications for the battery chargers were reviewed to verify that a charger failure would be detected before the next surveillance.

b. Findings

No findings of significance were identified.

.2.17 Reactor Vessel Level Instrumentation Used During Shutdown/Mid-loop Operations

a. Inspection Scope

The team reviewed the reactor vessel level indication used during shutdown/mid-loop operations to ensure that level instrumentation and low level alarm setpoints had been properly incorporated. The team also reviewed the calculation that determined the accuracy for the Mansell Level Monitor System, which included the calibration procedures for the various sensing and signal processing components that were installed in the system, to verify that instrument uncertainty had been included. The team observed the use of the Mansell level monitors during a simulator exercise for mid-loop operations.

b. Findings

No findings of significance were identified.

.2.18 Emergency Diesel Generator (1A-A) Load Sequencing

a. Inspection Scope

The team reviewed schematic diagrams for emergency load sequencing for design basis conditions. The team reviewed the corrective maintenance history for the sequencing relays for problems related to the failure of the relays to function on demand. The team also reviewed modification history to identify any changes and their effect on the design basis for equipment. The team reviewed load sequencing testing to verify that accident conditions were properly simulated and that the equipment response was is in accordance with design requirements.

b. Findings

No findings of significance were identified.

.3 Review of Low Margin Operator Actions

a. Inspection Scope

The team performed a margin assessment and detailed review of five risk significant and time critical operator actions. Where possible, margins were determined by the review of the assumed design basis and UFSAR response times and performance times documented by job performance measures (JPM) results. For the selected components and operator actions, the team performed an assessment of the Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs), Annunciator Response Procedures, and other operations procedures to determine the adequacy of the procedures and availability of equipment required to complete the actions. Operator actions were observed on the plant simulator and during plant walk downs for the following:

• Loss of RHR (mid loop operations)

• Manual local operation of TDAFW pump (with and without battery power available)

• Align ECCS for high pressure recirculation (autoswap successful)

• RHRP suctions to sump given auto swap fails

• Start standby ERCW pump on loss of running pump Additionally, the team reviewed the availability and adequacy of instrumentation used by operators in the above activities to verify appropriate indication was available to monitor plant conditions and plant response to operator actions.

b. Findings

LOSP Impact on Mansell level System

Introduction:

The team identified a violation of TS 6.8.1, related to the failure to establish an adequate procedure for reactor coolant system reduced inventory/mid-loop operations. Specifically, procedure 0-GO-13, Reactor Coolant System Drain and Fill Operations, Rev. 57, was not adequate in that it did not establish adequate actions to maintain continuous RCS level indication during all possible plant conditions while in the reduced inventory/mid-loop configuration, specifically loss of off site power (LOSP).

Additionally, the procedure did not establish contingency actions to recover power to the Mansell level indication systems or provide guidance for alignment of an alternate RCS level indication mechanism within the 30 minutes for which power was available from the Mansell system unit battery.

Description:

Procedure 0-GO-13, Reactor Coolant System Drain and Fill Operations, Rev. 57, implements the licensees commitment to GL 88-17, Loss of Decay Heat Removal (non power operations) and related FSAR section 5.6 requirements state that at least two independent continuous RCS level monitoring indications will be maintained during mid-loop operations. Per this procedure, two channels of the Mansell System provide the two continuous independent level monitoring systems required for reduced inventory/mid-loop operations. The instruments are powered from 120 VAC wall receptacles in the main control room which are powered by the auxiliary building lighting boards (LC-132 and LC-232, powered from Units 1 and 2 respectively). On a loss of off-site power the lighting boards would be de-energized and are not re-energized when the EDGs restore on-site vital power. Each Mansell unit has an uninterruptible power supply (UPS) battery to maintain the system for approximately 30 minutes. This was verified by on-site testing performed during the inspection. No procedure contingency plan exists for maintaining/restoring Mansell indication once the UPS is depleted.

Analysis:

The failure to establish an adequate procedure for reactor coolant system reduced inventory/mid-loop operations is a performance deficiency associated with the initiating events cornerstone. This finding is more than minor because it impacts the Cornerstones objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and is associated with the cornerstone attribute of procedure quality. This finding was reviewed for cross-cutting aspects and none were identified.

Enforcement:

TS 6.8.1 requires that procedures be established, implemented, and maintained covering the activities specified in Appendix A of Regulatory Guide 1.33, Revision 2. Paragraph 9d of the appendix requires procedures for draining and refilling of the reactor vessel. Contrary to the above, procedure 0-GO-13, Reactor Coolant System Drain and Fill Operations, Rev. 57, was not adequate in that it did not establish adequate actions to maintain continuous RCS level indication during all possible plant conditions while in the reduced inventory/mid-loop configuration, specifically loss of off site power. Additionally, the procedure did not establish contingency actions to recover power to the Mansell level indication systems or provide guidance for alignment of an alternate RCS level indication mechanism within the 30 minutes for which power was still available from the Mansell system unit battery. This finding was entered into the Sequoyah corrective action program as PER 125906. Because this finding is of very low safety significance and was entered into SNPs corrective action program, this violation is being treated as a non-cited violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000327,05000328/2007006-03, Inadequate Procedure for Reduced Inventory/Mid-loop Operation)

Procedure AOP-R.03, RHR System Malfunction

Introduction:

The team identified a violation of TS 6.8.1, related to the failure to establish an adequate abnormal operating procedure for RHR system malfunctions during shutdown conditions. Procedure AOP-R.03, RHR System Malfunction, Rev. 17, was not adequate in that it did not establish adequate actions to restore RHR cooling following isolation of an RHR leak during hot shutdown (Mode 4) operations. The instruction provided in the procedure could result in a total loss of RHR cooling capability during mode 4 conditions if an RHR leak occurred.

Description:

Procedure AOP-R.03, RHR System Malfunction, Rev. 17, provided instruction for protection of the reactor core during shutdown (non power) conditions (Modes 4,5 and 6) in the event of a loss of RHR cooling, RHR system leak, or a loss of RHR level. Section 2.4 provides instruction for mitigation of an RHR system leak. Step 2 of section 2.4 stated: if the magnitude of a leak requires rapid isolation, secure RHR pumps and close hot leg and individual RHR pump suction valves [valves 74-1, 74-2, 74-3, and 74-21]. (Note: magnitude of leak was not defined) Steps 3 through 9 involved identifying and isolating the RHR leak. Step 10 directed the operator to place the unaffected RHR loop in service using procedure 0-SO-74-1, System Operating Procedure for RHR System. Entry points into this procedure would be at section 5.5.2, Placing RHR in Service for Normal Shutdown Cooling or section 8.2, Swapping RHR Pumps with RCS in Mid-Loop Conditions. Section 5.5.2, directed the loop suction valves [74-1 and 74-2] to be opened; however, there was no direction to open or verify open the individual pump suction valves [74-3 and 74-21] before the direction to start the pumps nor did section 8.2. The team concluded the pumps could be started without a suction flow path, resulting in pump damage in a relatively short period (minutes),while running on mini flow with the pump suction valves closed due to a lack of NPSHA.

Analysis:

The failure to establish an adequate abnormal operating procedure for RHR system malfunctions during shutdown conditions is a performance deficiency associated with the Initiating Events Cornerstone. This finding is more than minor because it impacts the Cornerstones objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown in that the loss of RHR pumps would increase the likelihood of a loss of RHR cooling. This finding was reviewed for cross-cutting aspects and none were identified.

Enforcement:

TS 6.8.1 requires that procedures be established, implemented, and maintained covering the activities specified in Appendix A of Regulatory Guide 1.33, Revision 2. Paragraph 5 of the of the appendix requires procedures for abnormal, off normal and alarm conditions. Contrary to the above, procedure AOP-R.03, RHR System Malfunction, Rev. 17, was not adequate in that it did not establish adequate actions to restore RHR cooling following isolation of an RHR leak which could occur during Mode 4 operations. This finding was entered into the Sequoyah corrective action program as PER 125844 with corrective actions taken to revise procedure AOP-R.03 to assure the pump suction valves are open before the transition to procedure 0-SO-74-1 to restore RHR cooling. Because this finding is of very low safety significance and was entered into SNPs corrective action program, this violation is being treated as a non-cited violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000327,05000328/2007006-04, Inadequate Abnormal Operating Procedure for RHR System Malfunctions During Mode 4 Conditions)

.4 Review of Industry Operating Experience

a. Inspection Scope

The team reviewed selected operating experience issues that had occurred at domestic and foreign nuclear facilities for applicability at the Sequoyah Nuclear Plant. The team performed an independent applicability review and issues that appeared to be applicable to the Sequoyah Nuclear Plant were selected for a detailed review. The issues that received a detailed review by the team included:

IN 2002-012, Submerged Safety-Related Electrical Cables IN 2006-17, Recent operating experience of service water systems due to external conditions.

IN 2007-05, Vertical deep draft pump shaft and coupling failures.

IN 2007-06, Potential common cause vulnerabilities in essential service water systems.

GL 88-17, Loss of decay heat removal.

b. Findings

No findings of significance were identified.

.5 Review of Permanent Plant Modifications

a. Inspection Scope

The team reviewed three modifications related to the selected risk significant components in detail to verify that the design bases, licensing bases, and performance capability of the components have not been degraded through modifications. The adequacy of design and post modification testing of these modifications was reviewed by performing activities identified in IP 71111.17, Permanent Plant Modifications, Section 02.02.a. Additionally, the team reviewed the modifications in accordance IP 71111.02, Evaluations of Changes, Tests, or Experiments, to verify the licensee had appropriately evaluated them for 10 CFR 50.59 applicability. The following modifications were reviewed:

• DCN D-22161A, Install ERCW High Point Vents on ERCW Discharge Headers, Rev. A

• DCN D-21550, AFW Level Control Valves-Actuator Capability Increase, Rev. A

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

4AO6 Meetings On July 13, 2007, the team presented the preliminary inspection results to Mr. Pitesa, Station Manager, and other members of the licensee staff. On September 10, 2007 the team presented the final inspection results to members of the licensee staff. The team returned all proprietary information examined to the licensee. No proprietary information is documented in the report.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

M. Casner, Inspection Support Manager

K. Nesmith, Inspection Team Leader

N. Thomas, Regulatory Affairs Lead

D. Porter, Op Time Critical Actions

C. Wilson, Equipment Manager

G. Bell, Electrical Design Engineer

D. Dimopoulos, Electrical Engineer

E. Craig, Mechanical System Engineer

S. Carter, Mechanical Design Engineer

NRC

B. Desai, RII, Branch Chief, Engineering Branch 1

S. Freeman, Senior Resident Inspector

M. Speck, Resident Inspector

ITEMS OPENED, CLOSED, AND DISCUSSED

Open/Closed

05000327,328/2007006-01 NCV Failure To Use Appropriate Assumptions in Design Calculations. (Section 1R21.2.2)

05000327,328/2007006-02 NCV No Procedure for Inspection/Replacement of DG Local Electrical Panel Filters. (Section 1R21.2.13)

05000327,328/2007006-03 NCV Inadequate Procedure for Reduced Inventory/Mid-loop Operation. (Section 1R21.3)

05000327,328/2007006-04 NCV Inadequate Abnormal Operating Procedure for RHR System Malfunctions During Mode 4 Conditions. (Section 1R21.3)