IR 05000348/2011010

IR 05000348-11-010, 05000364-11-010; on 08/29/2011 - 12/08/2011; Joseph M. Farley Nuclear Plant, Units 1 and 2; Component Design Bases Inspection

ML113530575
Person / Time
Site:	Farley
Issue date:	12/19/2011
From:	Nease R NRC/RGN-II/DRS/EB1
To:	Lynch T Southern Nuclear Operating Co
References
IR-11-010
Download: ML113530575 (49)

Text

December 19, 2011

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT - NRC COMPONENT DESIGN BASES INSPECTION - INSPECTION REPORT 05000348/2011010 AND 05000364/2011010

Dear Mr. Lynch:

On, December 8, 2011, U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Joseph M. Farley Nuclear Plant, Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on December 8, 2011, with Mr. Todd Youngblood 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 licenses. The team reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents six NRC identified findings of very low safety significance (Green), which were determined to involve violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent the NRC Enforcement Policy. If you contest these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-001; with copies to the Regional Administrator Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Farley. Further, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at Farley. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

In accordance with 10 CFR 2.390 of the NRC's "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 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 by Binoy B. Desai For/

Rebecca Nease, Chief Engineering Branch 1 Division of Reactor Safety

Enclosure:

Inspection Report 05000348, 364/2011010,

w/Attachment: Supplemental Information

Docket No.: 50-348, 50-364 License No.: NPF-2, NPF-8

REGION II==

Docket Nos.: 050000348, 05000364

License Nos.: NPF-2, NPF-8

Report Nos.: 05000348/2011010 and 05000364/2011010

Licensee:

Southern Nuclear Operating Company, Inc.

Facility:

Joseph M. Farley Nuclear Plant, Units 1 and 2

Location:

Columbia, AL

Dates:

August 29 - December 8, 2011

Inspectors:

S. Sandal, Senior Reactor Inspector (Lead)

D. Jones, Senior Reactor Inspector J. Eargle, Reactor Inspector D. Mas Penaranda, Reactor Inspector M. Shlyamberg, Accompanying Personnel P. Wagner, Accompanying Personnel

Approved by: Rebecca Nease, Chief

Engineering Branch 1

Division of Reactor Safety

Enclosure

SUMMARY OF FINDINGS

IR 05000348/2011-010, 05000364/2011-010; 08/29/2011 - 12/08/2011; Joseph M. Farley

Nuclear Plant, Units 1 and 2; Component Design Bases Inspection.

This inspection was conducted by a team of four NRC inspectors from the Region II office, and two NRC contract personnel. Six Green non-cited violations (NCV) were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using the NRC Inspection Manual Chapter (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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, (ROP)

Revision 4, dated December 2006.

NRC identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

The team identified a non-cited violation (NCV) of 10 CFR 50, Appendix B,

Criterion III, Design Control, (with two examples) for the licensees failure to implement design control measures to verify the adequacy of design inputs, assumptions, or limiting plant conditions which were relied upon in the design basis analyses used to demonstrate the adequacy of Condensate Storage Tank (CST)design. The licensee entered these issues into their Corrective Action Program (CAP) as Condition Reports (CRs) 355226, 355293, and 355294. The licensee performed operability evaluations in support of current operability and implemented additional compensatory measures to ensure that CST level would be maintained above the condenser hotwell make-up elevation pending completion of proposed long term corrective actions which included a license amendment request to increase the minimum volume of water specified by the limiting condition for operation in Technical Specification (TS) 3.7.6.

The failure to utilize conservative design inputs, assumptions, or limiting plant conditions when implementing design control measures to verify the adequacy of CST design was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the mitigating systems cornerstone attribute of design control and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding challenged the assurance that the CST contained an adequate volume of water to support its safety function to supply condensate to the Auxiliary Feedwater (AFW) system in response to design basis events. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 Significance Determination Process (SDP) screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. This analysis was based on information contained in licensee operability determinations which demonstrated that, although the TS required minimum volume of 150,000 gallons was non-conservative, reasonable assurance existed such that the volume of CST water below the condenser hotwell make-up elevation was sufficient for the tank to perform its safety function. A cross-cutting aspect was not identified because the design basis calculation associated with the finding was approved on March 25, 1999, and did not represent current licensee performance. [Section 1R21.2.3]

Green.

The team identified a non-cited violation (NCV) of 10 CFR 50, Appendix B,

Criterion III, Design Control, for the failure to implement design control measures to verify the adequacy of design inputs, assumptions, or limiting plant conditions which were relied upon in the design basis analyses used to demonstrate the capability of the Auxiliary Feedwater (AFW) system to deliver the required flowrates to the Steam Generators (SGs). The licensee entered this issue into the Corrective Action Program (CAP) as Condition Reports (CRs) 352210, 353743, 355898, 363850, and 369676. Additionally, the licensee performed an operability determination which concluded that the AFW system remained capable of performing its safety function because actual AFW pump performance was not degraded as assumed in the accident analyses.

The failure to conservatively model AFW system friction losses when implementing design control measures to verify the capability of the AFW system to deliver the flowrates required by accident analyses was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the mitigating systems cornerstone attribute of design control and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding challenged the assurance that the AFW system would be capable of delivering the required flow during worst case accident conditions due to non-conservative modeling of system friction losses. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 Significance Determination Process (SDP) screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification (TS) allowed outage time, and did not affect external event mitigation. A cross-cutting aspect was not identified because the design basis calculation associated with the finding was approved on March 25, 1999, and did not represent current licensee performance. [Section 1R21.2.3]

Green.

The team identified a non-cited violation (NCV) of Technical Specification (TS) 5.4, Procedures, for the licensees failure to provide adequate procedural guidance for controlling steam generator (SG) and pressurizer level during loss of instrument air events and Chemical and Volume Control System (CVCS)malfunctions. Specifically, the licensee failed to evaluate the capability of motor-operated valves (MOVs) to be cycled as directed by abnormal operating procedures (AOPs). The licensee entered these issues into their Corrective Action Program (CAP) as Condition Reports (CRs) 355230, 355672 and 355695; performed DOEJ -

FRSNC326893-E001, Evaluate Cycling of Q1E21MOV8107, Q1E21MOV8107, and Q1E21MOV3764A through F; and implemented a standing order (S-2011-12) that restricted the cycling the of the MOVs until the procedures were revised.

The failure to provide adequate procedural guidance for controlling SG and pressurizer level during loss of air events and CVCS malfunctions was a performance deficiency. The performance deficiency was more that minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee directed the cycling of MOVs in AOPs without performing evaluations to provide assurance that the components would not fail as a result of the cycling operations and lead to a condition of inadequate SG and pressurizer level control. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 Significance Determination Process (SDP) screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification (TS) allowed outage time, and did not affect external event mitigation. A cross-cutting aspect was not identified because the finding did not represent current performance. [Section 1R21.2.3]

Green.

The team identified a non-cited violation (NCV) of 10 CFR 50, Appendix B,

Criterion III, Design Control, involving two examples. In the first example, the licensee failed to translate the minimum Component Cooling Water (CCW) flow for the Residual Heat Removal (RHR) seal coolers into Annunciator Response Procedures (ARPs). In the second example, the licensee failed to translate the Motor Driven Auxiliary Feedwater (MDAFW) and Turbine Driven Auxiliary Feedwater (TDAFW) pump minimum flow requirements into applicable ARPs. The licensee entered these issues into their Corrective Action Program (CAP) as Condition Reports (CRs) 348613 and 352485.

The failure to correctly translate the applicable design bases information for the RHR pump seal coolers and the Auxiliary Feedwater (AFW) pumps into procedures was a performance deficiency. The finding was determined to be more than minor because it was associated with the procedure quality attribute of the mitigating system cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to translate the appropriate minimum flow requirements into ARPs adversely affected the quality of procedures used to respond to alarm conditions that are required by Regulatory Guide 1.33,

Quality Assurance Program Requirements. The inadequate procedures adversely affected the ability of operators to assess operability and to combat deficiencies associated with risk significant equipment. In accordance with NRC IMC 0609.04,

Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 Significance Determination Process (SDP)screening and determined the finding to be of very low safety significance (Green)because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification (TS) allowed outage time, and did not affect external event mitigation.

A cross-cutting aspect was not identified because the finding did not represent current performance. [Section 1R21.2.4]

Green.

The team identified a non-cited violation (NCV) of 10 CFR 50.65,

Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to perform condition monitoring or otherwise implement an appropriate preventive maintenance program for the 2C Diesel Generator (DG) A and B room exhaust fan louvers. The licensee entered this issue into their corrective action program (CAP) as condition reports (CRs) 351580, 349883, and 355130.

The failure to perform condition monitoring or otherwise implement an appropriate preventive maintenance program for the 2C DG A and B exhaust fan louvers was a performance deficiency. This performance deficiency was more than minor because it was associated with equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform condition monitoring or otherwise implement an appropriate preventive maintenance program for the 2C DG A and B room exhaust fan louvers challenged the assurance that these components would remain capable of performing their intended functions. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 SDP screening and determined the finding to be of very low safety significance (Green)because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. Because the licensee had initiated CRs in 2008 and 2009 for the 2C DG room exhaust louvers, and repairs were not made in a timely manner to address the issue, this finding was assigned a cross-cutting aspect in the corrective action program component of the problem identification and resolution area P.1(d). [Section 1R21.2.6]

Green.

The team identified a non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish an adequate test procedure used to demonstrate that the Turbine Driven Auxiliary Feedwater (TDAFW) pump discharge check valves were capable of performing their design basis function. The test procedure was inadequate in providing assurance that the Auxiliary Feedwater (AFW) system was capable of providing the required design basis flow rates to the Steam Generators (SGs) with reverse flow into an idle TDAFW pump via the discharge check valves. This issue was entered into the licensees Corrective Action Program (CAP) as Condition Report (CR) 348795.

The failure to develop an adequate test procedure which demonstrated that TDAFW pump discharge check valves were capable of performing their design basis function was a performance deficiency. This performance deficiency was more than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the acceptance criteria used in the test procedure was non-conservative when compared to the flow rates required by the accident analyses, and the test procedure was performed at lower system pressures (which were not representative of actual design conditions). In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 Significance Determination Process (SDP) screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification (TS)allowed outage time, and did not affect external event mitigation. Because the test procedure did not contain complete, accurate, and up-to-date information consistent with the system design basis safety analysis, this finding is assigned a cross-cutting aspect in the resources component of the human performance area H.2(c). [Section 1R21.2.7]

Licensee-Identified Violations

None

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R21 Component Design Bases Inspection

.1 Inspection Sample Selection Process

The team selected risk significant components and related 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 1.3 or Birnbaum value greater than 1 X10-6. The sample included fifteen components, including one associated with containment large early release frequency (LERF), and four operating experience (OE) items.

The team performed a margin assessment and a detailed review of the selected risk-significant components to verify that the design bases had been correctly implemented and maintained. This margin assessment considered original design issues, margin reductions due to modifications, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for a detailed review. These reliability issues included items related to failed performance test results, significant corrective action, repeated maintenance, maintenance rule status, Regulatory Issue Summary (RIS)05-020 (formerly Generic Letter (GL) 91-18) conditions, NRC resident inspector input of problem equipment, System Health Reports, industry OE, 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 is included in the following sections of the report.

.2.1 Steam Generator (SG) Main Steam Isolation Valves (MSIV) - Q2N11HV3369(3370)

a. Inspection Scope

The team reviewed the updated Final Safety Analysis Report (UFSAR), TS, Functional System Description (FSD), and piping and instrumentation diagrams (P&IDs), applicable plant calculations, and drawings to identify the design bases requirements of the MSIVs.

The team examined system health reports, records of surveillance testing and maintenance activities, and applicable corrective actions to verify that potential degradation or low margin design issues were being monitored, prevented and/or corrected. Additionally, the team reviewed station operating and off-normal response procedures to verify design bases requirements had been adequately translated into procedural instructions. The team performed a walkdown of the valve areas. The team reviewed design bases documentation, maintenance records, and drawings of the instrument air system to verify that the support function provided to the MSIVs was consistent with design requirements. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions.

b. Findings

No findings were identified.

.2.2 Steam Generator Atmospheric Relief Valves (ARVs) - Q1N11PV3371A(B/C) - LERF

a. Inspection Scope

The team reviewed the UFSAR, TS, FSD, P&IDs, applicable plant calculations, and drawings to identify the design bases requirements of the ARVs. The team examined system health reports, records of surveillance testing and maintenance activities, and applicable corrective actions to verify that potential degradation or low margin design issues were being monitored, prevented and/or corrected. Additionally, the team reviewed station operating and off-normal response procedures to verify design bases requirements had been adequately translated into procedural instructions. The team performed a walkdown of the valve areas. The team reviewed design bases documentation, maintenance records, and drawings of the instrument air system to verify that the support function provided to the ARVs was consistent with design requirements. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. The team reviewed off-normal and emergency operating procedures to verify that adequate guidance exists for operators to respond to a design bases steam generator tube rupture event. The team observed a simulator scenario to verify the capability of the operators to mitigate a Steam Generator Tube Rupture (SGTR) event as described in the UFSAR. The team performed a walkdown of local manual actions associated with a SGTR event to verify the feasibility of the directed actions.

b. Findings

No findings were identified.

.2.3 Motor Driven Auxiliary Feedwater (MDAFW) Pump (Mechanical) - Q2N23P001B

a. Inspection Scope

The team reviewed the plants TS, UFSAR, FSD, and P&IDs to establish an overall understanding of the design bases of the MDAFW. Design calculations and test data were reviewed to verify that design basis capability, and flow rates had been appropriately translated into these documents. The team concentrated its efforts on the pumps capability of performing its safety function (i.e., delivering the required flow rate to the steam generators at the prescribed design pressure). Records of surveillance testing and maintenance activities, and applicable corrective actions were examined to verify that potential degradation or low margin design issues were being monitored, prevented and/or corrected. MDAFW walkdowns were conducted to verify that the installed configurations would support its design basis function under accident conditions and had been maintained to be consistent with design assumptions and to visually inspect the material condition of the pumps. Control panel indicators were observed and operating procedures reviewed to verify that MDAFW operation and alignments were consistent with design and licensing basis assumptions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. The team reviewed off-normal procedures to verify that adequate guidance exists for operators to control pressurizer and SG levels during loss of air events. The team performed a walkdown of local manual actions associated with controlling pressurizer and SG levels to verify the feasibility of the directed actions.

b.

.1 Findings

Introduction:

A Green NRC identified NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, (with two examples) was identified for the failure to implement design control measures to verify the adequacy of design inputs, assumptions, or limiting plant conditions which were relied upon in the design basis analyses used to demonstrate the adequacy of Condensate Storage Tank (CST) design.

Description:

The CST is a safety related, seismic category I tank that holds up to 500,000 gallons of water and is required by the TS 3.7.6 limiting condition for operation to be maintained at a minimum of 150,000 gallons for use by the AFW system under normal operation and in response to accident conditions. In order to ensure this requirement, the lower 13 3-1/8 of the 46 inside diameter (ID) tank is designed to withstand the effects of tornado missiles. The CST has two 8 AFW suction pipes - one for the TDAFW pump and one for both MDAFW pumps. Both suction pipes open at 4 from the tanks bottom facing down. The suction piping centers are approximately 1 3 apart. The CST has an internal bladder that prevents introduction of air under normal operating conditions.

The team reviewed calculations BM-95-0961-001, Verification of CST Sizing Basis, Rev. 4, dated March 25, 1999, and CBI-72-4859, Condensate Storage Tank, Rev. 0 and identified two examples where the licensee had failed to implement design control measures to verify the adequacy of design inputs, assumptions, or limiting plant conditions which were relied upon in design basis analyses used to demonstrate that the CST would have a sufficient volume of water to perform its safety function. The following examples were identified:

Example 1 - Effect of +2% Calorimetric Error and 15 Megawatts Thermal (MWt) Heat Input from the Reactor Coolant Pumps (RCPs) - The team reviewed the applicable design basis analysis for the CST and noted that calculation BM-95-0961-001 used non-conservative design inputs and did not take into consideration +2% calorimetric error when establishing the initial assumed reactor thermal power as discussed in UFSAR accident analysis. Additionally, the team noted that the calculation used 10 MWt and not 15 MWt as a heat input from the RCPs as also discussed in UFSAR accident analysis.

The team concluded that the use of non-conservative inputs in the CST design analyses adversely impacted the margin available in the TS 3.7.6 required CST volume. Based on the teams observations, the licensee entered the issue into the CAP as CR 355226 and documented their operability determination in PDO 0-11-06. The licensees evaluation demonstrated that these non-conservative assumptions resulted in an increase in required water volume by 1,648 gallons in excess of the value originally calculated in BM-95-0961-001. The operability determination also concluded that sufficient water remained available below the condenser hotwell make-up elevation for the CST to be able to perform its safety function.

Example 2 - Effect of Tavg Assumed in the Accident Analysis on CST Volume Requirements - The team reviewed the applicable design basis analysis for the CST and noted that calculation BM-95-0961-001 did not address the CST volume requirements for the Main Feed Line Break (MFLB) case that TS Bases 3.7.6 described as the limiting event for the required condensate volume. The licensee entered this issue into the CAP as CR 355293. The licensees evaluation of this observation identified that the volume required in the MFLB case was 416 gallons less than the loss of offsite power cooldown case volume. However, the team also noted that the licensees evaluation was based on the Reactor Coolant System (RCS) sensible heat calculated for the no-load case where RCS Tavg is nominally 547 °F. As documented in UFSAR accident analysis, RCS Tavg for the MFLB at time zero was assumed to be higher than 547 °F at approximately 583 °F. Additionally, the team noted that UFSAR accident analysis assumed a vessel average temperature as high as 577.2 °F and stated that +/- 6 °F steady-state Tavg error was considered in the analysis. From this information, the team concluded that the nominal use of 547 °F Tavg in the CST design basis analysis was not consistent with the assumptions stated in UFSAR accident analysis and was non-conservative in determining the most limiting CST required water volume. The teams evaluation of the effect of the increased Tavg on the required CST volume could be as high as an additional 5,565 gallons. The team concluded that the use of non-conservative Tavg as an input in the CST design basis analysis for RCS sensible heat load adversely impacted the margin available in the TS 3.7.6 required CST volume. Based on the teams observations, the licensee entered the issue into the CAP as CR 355294 to address the sensible heat concern.

The team reviewed applicable operability determinations completed by the licensee regarding the issues identified above and concluded that although the TS required minimum volume of 150,000 gallons was determined to be non-conservative, reasonable assurance existed such that the volume of CST water below the condenser hotwell make-up elevation was sufficient for the tank to remain capable of performing its safety function (at reduced margin). As a result of the teams observations, the licensee implemented additional compensatory measures to ensure that CST level would be maintained above the condenser hotwell make-up elevation pending completion of proposed long term corrective actions (including a license amendment to modify the TS required minimum volume). The team also reviewed longer term proposed licensee corrective actions to revise the applicable design basis calculations for the CST.

Analysis:

The failure to utilize conservative design inputs, assumptions, or limiting plant conditions when implementing design control measures to verify the adequacy of CST design was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the mitigating systems cornerstone attribute of design control and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding challenged the assurance that the CST contained an adequate volume of water to support its safety function to supply condensate to the AFW system in response to design basis events. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 SDP screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. This analysis was based on information contained in licensee operability determinations which demonstrated that, although the TS required minimum volume of 150,000 gallons was non-conservative, reasonable assurance existed such that the volume of CST water below the condenser hotwell make-up elevation was sufficient for the tank to perform its safety function. A cross-cutting aspect was not identified because the design basis calculation associated with the finding was approved on March 25, 1999, and did not represent current licensee performance.

Enforcement:

10 CFR 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall provide for verifying or checking the adequacy of design. Contrary to the above, since March 25, 1999, the licensee failed to implement design control measures as described in the two examples above to verify the adequacy of design inputs, assumptions, or limiting plant conditions which were relied upon in design basis calculations used to demonstrate the adequacy of CST design. Because the violation was of very low safety significance and it was entered into the licensees CAP, this violation is being treated as an NCV consistent with the NRC Enforcement Policy: NCV 05000348, 364/2011010-01, Failure to Implement Design Control Measures to Verify the Adequacy of CST Design.

b.

.2 Findings

Introduction:

A Green NRC identified NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the failure to implement design control measures to verify the adequacy of design inputs, assumptions, or limiting plant conditions which were relied upon in the design basis analyses used to demonstrate the capability of the AFW system to deliver the required flowrates to the Steam Generators (SGs).

Description:

The AFW system is a safety related, seismic category I system that is credited to provide the required cooling water from the CST to each of the three Steam Generators (SGs). The AFW system design provides for redundancy by assuring that either a single TDAFW pump or two MDAFW pumps can deliver the required flows. The licensee established acceptability of the design based, in part, on the results of Unit 1 calculation 40.02, Verification of AFW Flow Bases, Rev. 4, Unit 2 calculation 38.04, Verification of AFW Flow Bases, Rev. 4, and calculation 11.13, Available NPSH for Auxiliary Feedwater Pumps, Rev. 1. The AFW flow bases calculations identified a main feed water line break (MFLB) accident with the failure of the TDAFW pump as the most limiting case in terms of design flow margin to the other two steam generators.

Calculation 40.02 identifies a limiting flow margin of 2.21 gallons per minute (gpm) per SG. Calculation 38.4 identifies a limiting flow margin of 0.82 gpm per SG. These margins are the flowrates available above the accident analysis required flowrate of 150 gpm.

The team reviewed the applicable design basis analyses for the MDAFW pumps and noted that calculations 40.02 and 38.04 used a non-conservative assumption in modeling friction losses of the AFW flow orifices (FO) 2861 A, B, C and FO 2862 A, B, C. The orifice resistance modeling did not take into account variation of the orifice and/or piping inside diameter (ID). Additionally, the analysis did not consider the effects of the change in elevation and resistance of piping internal to the steam generators and did not use the fluid temperature that would maximize friction losses. Based on the teams observations, the licensee entered this issue into the CAP as CR 352210, CR 353743, CR 355898, CR 363850, and CR 369676 and performed an operability determination to address the impact of the non-conservative assumptions on AFW design analyses. DJ-FRSN2689-005, Documentation of Engineering Judgment, Hydraulic Evaluation of FNP Auxiliary Feedwater System to Support Operability Determinations, Ver. 1.0, determined that the orifice resistance is approximately 70% of the total AFW system resistance. Therefore, modeling of orifice resistance is critical to the validity of the AFW analysis. The licensees evaluation demonstrated that once the model was corrected to account for these non-conservative assumptions, the predicted flow to the non-affected SGs for the MFLB case would have been below the required design basis flowrate of 150 gpm (assuming 5% degraded pump performance). The operability determination concluded that the AFW system remained operable because actual pump performance was not degraded, and therefore, would yield flows in excess of the required 150 gpm.

Analysis:

The failure to conservatively model AFW system friction losses when implementing design control measures to verify the capability of the AFW system to deliver the flowrates required by accident analyses was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the mitigating systems cornerstone attribute of design control and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding challenged the assurance that the AFW system would be capable of delivering the required flow during worst case accident conditions due to non-conservative modeling of system friction losses. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 SDP screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. A cross-cutting aspect was not identified because the design basis calculation associated with the finding was approved on March 25, 1999, and did not represent current licensee performance.

Enforcement:

10 CFR 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall provide for verifying or checking the adequacy of design. Contrary to the above, since March 25, 1999, the licensee had failed to implement design control measures to verify the adequacy of design inputs, assumptions, or limiting plant conditions which were relied upon in design basis calculations used to demonstrate the adequacy of AFW design to meet flowrates required by the accident analysis. Because the violation was of very low safety significance and it was entered into the licensees CAP, this violation is being treated as an NCV consistent with the NRC Enforcement Policy: NCV 05000348, 364/2011010-02, Failure to Implement Design Control Measures to Verify the Adequacy of AFW Design.

b.

.3 Findings

Introduction:

A Green, NRC identified NCV of TS 5.4, Procedures, was identified for the licensees failure to provide adequate procedural guidance for controlling SG and pressurizer level during loss of instrument air events and CVCS malfunctions.

Specifically, the licensee failed to evaluate the capability of MOVs to be cycled as directed by AOPs.

Description:

Pressurizer level and SG levels are normally controlled by air-operated flow control valves (FCVs). During malfunctions, such as a loss of instrument air, the FCVs fail in the full open position. To prevent overfill of the pressurizer and SGs with a failed open FCV, the operators are directed by procedure to cycle (close - open - close)normally open motor operated valves (MOV) in the respective flow paths to control levels.

The licensee uses procedure FNP-1-AOP-6.0, Loss of Instrument Air, to mitigate a loss of instrument air event. Step 7 of the procedure directs operators to cycle MOV 8107 or 8108 to maintain pressurizer level between 20 to 50 percent. Step 8 of the procedure directs operators to cycle MOVs 3764A - F to maintain SG narrow range levels between 35 to 69 percent. The licensee uses procedure FNP-1-AOP-16.0, Chemical and Volume Control System (CVCS) Malfunction, to mitigate malfunctions of the charging and letdown portions of the CVCS. Step 22 of the procedure directs operators to cycle MOV 8107 or 8108 to maintain pressurizer level between 20 to 60 percent.

The team determined that the procedural guidance for controlling pressurizer and SG levels was inadequate because the licensee failed to evaluate the capability of the MOVs to be cycled as directed in the procedures. A MOV failure could occur as a result of the cycling due to the tripping of thermal overload devices or overheating of other electrical components. The failure of the MOVs would result in inadequate level control until local manual control was established by operators. During the inspection, the team verified the feasibility of the local manual actions. The licensee entered these issues into their CAP as CRs 355230, 355672 and 355695; performed DOEJ -FRSNC326893-E001, Evaluate Cycling of Q1E21MOV8107, Q1E21MOV8107, and Q1E21MOV3764A through F; and implemented a standing order (S-2011-12) that restricted the cycling of the MOVs until the procedures were revised. This issue is also applicable to Unit 2.

Analysis:

The failure to provide adequate procedural guidance for controlling SG and pressurizer level during loss of air events and CVCS malfunctions was a performance deficiency. The performance deficiency was more that minor because it was associated with the mitigating systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee directed the cycling of MOVs in AOPs without performing evaluations to provide assurance that the components would not fail as a result of the cycling operations and lead to a condition of inadequate SG and pressurizer level control. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 SDP screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. A cross-cutting aspect was not identified because the finding did not represent current performance.

Enforcement:

Technical Specification 5.4, Procedures, states, in part, that written procedures shall be established, implemented, and maintained covering the activities recommended in Appendix A of Regulatory Guide 1.33, Rev. 2, February 1978. The Regulatory Guide states in part that safety-related activities, such as combating emergencies and other significant events (i.e. loss of instrument air) should be covered by written procedures.

Contrary the above, since 1987, the licensee failed to maintain an adequate written procedure for combating emergencies and other significant events. Specifically, the licensee failed to provide adequate procedural guidance in procedures AOP-6.0, Loss of Instrument Air and AOP-16.0, CVCS Malfunction, for controlling SG and pressurizer level during design bases events. Because the violation was of very low safety significance and it was entered into the licensees CAP, this violation is being treated as an NCV consistent with the NRC Enforcement Policy: NCV 05000348, 364/2011010-03, Failure to Provide Adequate Procedural Guidance for Controlling Steam Generator and Pressurizer Level During Loss of Air Events.

b.

.4 Findings

Introduction:

The team identified an unresolved item (URI) regarding the licensees evaluation of the minimum required submergence for the AFW pumps given the potential for vortex formation in the CST.

Description:

The CST is a safety related, seismic category I tank that holds up to 500,000 gallons of water and is required by the TS 3.7.6 limiting condition for operation to be maintained at a minimum of 150,000 gallons for use by the AFW system under normal operation and in response to accident conditions. In order to ensure this requirement, the lower 13 3-1/8 of the 46 inside diameter (ID) tank is designed to withstand the effects of tornado missiles. The CST has two 8 AFW suction pipes - one for the TDAFW pump and one for both MDAFW pumps. Both suction pipes open at 4 from the tanks bottom facing down. The suction piping centers are approximately 1 3 apart. The CST has an internal bladder that prevents introduction of air under normal operating conditions.

The team reviewed calculations BM-95-0961-001, Verification of CST Sizing Basis, Rev. 4, and CBI-72-4859, Condensate Storage Tank, Rev. 0 and made the following observations regarding the design basis of the CST:

The CST tornado missile-protected height of 13 3-1/8 is based on the elevation of the 24 condenser hotwell make-up line. The hotwell make-up line is not designed to withstand a design basis seismic event or damage from tornado missiles. The team computed the maximum protected volume (including the unusable lower 4 of the tank) to be approximately 164,841 gallons. The team noted that this volume did not take into account any CST fabrication tolerances. Calculation BM-95-0961-001 established that there was a margin of 4,300 gallons with respect to the TS 3.7.6 requirements for the CST. Although this calculation addressed the losses of CST inventory due to the line break, it did not analyze that this line break would create an air introduction path under the CST bladder, allowing a vortex to form, and adversely affect the usable volume of water in the CST. Additionally, the team noted that calculation BM-95-0961-001 did not evaluate the effects of tornado missile damage to the un-protected portion of the CST.

Tornado missile damage to the tank could also create an air introduction path under the CST bladder which would allow a vortex to form.

Based on the teams observations, the licensee entered the issue into the CAP as CRs 351170, 353599 and 355457 and performed a prompt determination of operability (PDO)0-11-06, Prompt Determination of Operability, Rev. 2 which concluded that vortex formation could lead to an additional loss of required CST level of 5.8 or 6,021 gallons.

Although the additional water required to account for CST vortexing exceeded the TS minimum required value, the licensee concluded that sufficient water remained available below the condenser hotwell make-up elevation for the CST to be able to perform its safety function. Additionally, the licensee implemented administrative measures to ensure that CST level was maintained above the level determined to be required by the licensees evaluation. The PDO conclusions are supported, in part, by calculation SM-SNC335993-001, CST AFW Pump Suction - Submergence Analysis, Ver. 1.0. This calculation utilizes a methodology based on Akalank K. Jain, Air Entrainment in Radial Flow towards Intakes, ASCE Journal of Hydraulic Division, September 1978, to determine the minimum submergence water level in the tank to prevent vortexing.

Summary: The team determined that additional inspection and consultation with a vortexing subject matter expert at NRC headquarters would be warranted to evaluate the licensees application of the methodology used for determining minimum AFW pump submergence. Additionally, the team concluded that additional evaluation of minimum required AFW pump submergence would be necessary to determine if this issue resulted in a more than minor performance deficiency. (URI 05000348, 364/2011010-04, Evaluation of CST Vortex Effect on AFW Pump Minimum Submergence)

b.

.5 Findings

Introduction:

The team identified an URI regarding the use of non-conservative assumptions in design bases analyses used to demonstrate adequate available AFW pump net positive suction head (NPSH) and subsequent analysis of the impact of AFW system operation during a loss of instrument air event on available NPSH.

Description:

The AFW system is a safety-related, seismic category I system that is credited to provide the required cooling water from the CST to each of the three Steam Generators (SG). The AFW system design provides for motive force redundancy by assuring that either a single TDAFW pump, or 2 MDAFW pumps can deliver the required flows. The licensee established acceptability of the design based, in part, on results of Unit 1 calculation 40.02, Verification of AFW Flow Bases, Rev. 4, Unit 2 calculation 38.04, Verification of AFW Flow Bases, Rev. 4, and calculation 11.13, Available NPSH for Auxiliary Feedwater Pumps, Rev. 1. The AFW flow bases calculations identified a main feed water line break (MFLB) accident with the failure of the TDAFW pump as the most limiting case in terms of design flow margin to the other two steam generators.

Calculation 40.02 identifies a flow margin of 2.21 gpm per SG for SG A and B, and calculation 38.4 identifies a flow margin of 0.82 gpm per SG for SG A and C vs. the required flow of 150 gpm. Additionally, NPSH calculation 11.13 established that for a bounding case (2 MDAFW pumps operating) the margin between available NPSH (NPSHA) and required NPSH (NPSHR) was less than 1 foot.

The team reviewed calculations 40.02, 38.04 and 11.13 and identified examples where non-conservative design inputs, assumptions, or limiting plant conditions were relied upon in design basis analyses used to demonstrate that the MDAFW pumps would have a sufficient capacity and head to perform their safety function. The following specific examples were identified:

• Calculation 11.13 AFW flowrates were based on the flow rates developed in calculation 40.02 for the MSLB case with all three AFW pumps operating. However, for the NPSHA calculation, a conservative assumption would have been a failure of TDAFW pump, since it would maximize the flow through the remaining MDAFW pumps.

• NPSHA value was based on a CST temperature of 100°F and not 110°F as specified in FSAR Section 9.2.6.3 as the maximum CST temperature. In response to this observation, Farley performed an evaluation that established that use of 110 °F temperature resulted in decrease of NPSHA by 0.75 feet.

Based on the team observations, the licensee entered this issue into the CAP as CR 355025 and CR 352168 and performed an operability determination to address the impact of the non-conservative assumptions on AFW safety function. The licensees evaluation documented in IDO 355898 and DJ-FRSN2689-005 concluded that reasonable assurance existed that AFW safety function would not be adversely impacted by crediting operator actions to manually isolate the faulted SG after 30 minutes of operation (for MFLB) and by crediting the remaining CST level to add an approximately 9 foot increase in the NPSHA value in comparison to the one used in the calculation 11.13.

This CST level increase leads to the corresponding NPSHA greater than the NPSHR.

The teams review of IDO 355898 and DJ-FRSN2689-005 identified the following concerns and observations:

• DJ-FRSN2689-005 did not address NPSHA vs. NPSHR conditions at the lower CST levels. The team noted that abnormal operating procedures used to combat a loss of non-safety related instrument air allow cycling of AFW header MOVs in lieu of local-manual throttling of the AFW flow control valves. Since the pump flow rates will remain virtually the same every time the operators will open AFW isolation MOVs, the NPSHA advantage credited in the DJ-FRSN2689-005 will be decreasing until it will become negative prior to the CST becoming empty (since the 9 foot of added margin is less than the height of the 13 feet of protected volume).

• The licensees analyses did not address the potential for long-term AFW flow restricting orifice erosion that could lead to increased AFW flowrates and decreased NPSHA. Because the actual performance of the AFW flow restricting orifices is not periodically compared to the performance assumed in the design basis analyses, the team did not have sufficient information to conclude that the orifice bores would not erode undetected resulting in degraded performance.

Summary: The team determined that additional information and/or evaluation by the licensee were required to determine if operation of the AFW system during a loss of instrument air event (as described above) was consistent with system design basis assumptions and operability determinations. Additionally, the team determined that additional information from the licensee regarding the condition or performance of the AFW flow restricting orifices would be necessary to establish that current component performance remains consistent with design basis analyses and operability determinations. Additionally, the team concluded that these additional licensee evaluations would be necessary to determine if this issue resulted in a more than minor performance deficiency. (URI 05000348, 364/2011010-05, Non-Conservative Assumptions Regarding AFW Net Positive Suction Head)

.2.4 Residual Heat Removal (RHR) Pump Seal Coolers - Q1E11P001

a. Inspection Scope

The team reviewed the plants TS, UFSAR, Functional System Description (FSD), and piping and instrumentation diagrams (P&IDs) to establish an overall understanding of the design bases of the seal coolers. Design calculations were reviewed to verify that design basis heat removal requirements, capability, and flow rates had been appropriately translated into these documents. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.

Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Maintenance Rule (MR) information was reviewed to verify that the component was properly scoped, and that appropriate preventive maintenance was being performed to justify current MR status.

b. Findings

Introduction:

A green NRC identified NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified involving two examples. In the first example, the licensee failed to translate the minimum CCW flow for the RHR seal coolers into ARPs.

In the second example, the licensee failed to translate the MDAFW and TDAFW pump minimum flow requirements into applicable ARPs.

Description:

The team identified two examples of the licensee not translating the design of components into procedures required by Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation).

RHR Seal Cooler Design Flow Requirements - The RHR seal coolers utilize CCW flow to maintain the RHR pump seals below their design temperature limits. Low flow alarms are provided for the CCW flow on the return lines of the seal coolers. These alarms are received in the control room.

Calculation CN-96-0047, Component Cooling Water System Evaluation - Power Uprate and Replacement Steam Generator, Rev. 8 establishes that the minimum CCW flow to maintain the RHR Pump Seal Cooler process fluid outlet temperature below its maximum temperature of 180ºF is 3.5 gpm. The team noted that the licensee established low CCW flow alarm setpoints of 3 gpm (+1/ -0 gpm). These setpoints were then translated into the following Unit 1 and Unit 2 procedures: FNP-1(2)-ARP-1.3, Main Control Board Annunciator Panel C, Versions 28.1 and 22. The team determined that the setpoints in these procedures were non-conservative (low) when compared to calculated minimum design requirements and if left uncorrected could result in (1)inadequate CCW flow to the RHR seal coolers without the operators receiving the alarm in the control room, and

(2) the operators subsequently failing to start the standby train of RHR and secure the in-service train of RHR. The licensee entered this issue into their CAP as CR 348613.

AFW Pump Minimum Flow Requirements - Unit 1 and Unit 2 procedures FNP-1(2)-

ARP-1.9, Version 47 are the alarm response procedures for responding to MDAFW and TDAFW pump low suction flow conditions. The MDAFW pump alarm setpoint is 40 gpm

(+/- 4.5 gpm) and the TDAFW pump alarm setpoint is 80 gpm (+/- 4.5 gpm).

In 2005, the licensee revised the alarm response procedures. The revision added a note that states that the minimum flow requirements for the MDAFW pump was 50 gpm and the TDAFW pumps was 100 gpm. The minimum flow requirements were established by the manufacturer of the pumps. The team noted the associated alarm setpoints were non-conservative because the MDAFW pump alarm setpoint is 40 gpm (+/- 4.5 gpm)which is less than the minimum required value of 50 gpm; and the TDAFW pump alarm setpoint is 80 gpm (+/- 4.5 gpm) which is less than the minimum required value of 100 gpm. The licensee entered this issue into their CAP as CR 352485.

Analysis:

The failure to correctly translate the applicable design bases information for the RHR pump seal coolers and the AFW pumps into procedures was a performance deficiency. The finding was determined to be more than minor because it was associated with the procedure quality attribute of the mitigating system cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the failure to translate the appropriate minimum flow requirements into ARPs adversely affected the quality of procedures used to respond to alarm conditions that are required by Regulatory Guide 1.33, Quality Assurance Program Requirements.

The inadequate procedures adversely affected the ability of operators to assess operability and to combat deficiencies associated with risk significant equipment. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 SDP screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. A cross-cutting aspect was not identified because the finding did not represent current performance.

Enforcement:

10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that Measures shall be established to assure that the design basis is correctly translated into procedures. Contrary to the above, the licensee failed to correctly translate the applicable design bases information for the RHR pump seal coolers and AFW pumps into procedures. Specifically, since 2005 the licensee failed to translate the minimum CCW flow requirements for the RHR pump seal coolers and the minimum flow requirements for the AFW pumps into ARPs. Because the violation was of very low safety significance and it was entered into the licensees CAP, this violation is being treated as an NCV consistent with the NRC Enforcement Policy: NCV 05000348, 364/2011010-06, Failure to Correctly Translate the Design Basis into Procedures for Minimum CCW Flow to the RHR Seal Coolers and Minimum Flow Requirements for the AFW Pumps.

.2.5 RHR Inlet Isolation Motor Operated Valves - Q1E11MOV8701(8702)

a. Inspection Scope

The team reviewed the plant TS, UFSAR, FSD, and P&IDs to establish an overall understanding of the design bases of the valves. Design calculations (i.e., differential pressure and required torque/thrust) were reviewed to verify that the design basis and design assumptions had been appropriately translated into these documents. The team reviewed calculations for degraded voltage at the MOV terminals to ensure worst-case voltage was used in calculating available motor output torque when determining margin.

The team reviewed calculations that establish control circuit voltage drop and thermal overload sizing and testing to verify the capability of the valve to operate during design bases events. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. Test procedures and recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and analyses served to validate component operation under accident conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Maintenance Rule (MR) information was reviewed to verify that the component was properly scoped, and that appropriate preventive maintenance was being performed to justify current MR status.

b. Findings

No findings were identified.

.2.6 2C Station Blackout (SBO) Diesel Generator (DG) - QSR43A504

a. Inspection Scope

The team reviewed the plant TS, UFSAR, FSD, and P&IDs to establish an overall understanding of the design bases of the air start system, fuel oil storage tank, and ventilation system. Design calculations and site procedures were reviewed to verify the design bases and design assumptions had been appropriately translated into these documents. The team reviewed system modifications over the life of the component to verify that the subject modifications did not degrade the components performance capability and were appropriately incorporated into relevant drawings and procedures.

Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident/event conditions and had been maintained to be consistent with design assumptions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Maintenance Rule (MR) information was reviewed to verify that the component was properly scoped, and that appropriate preventive maintenance was being performed to justify current MR status.

b. Findings

Introduction:

A Green NRC identified NCV of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, was identified for the failure to perform condition monitoring or otherwise implement an effective preventive maintenance (PM) program for the 2C DG room A and B exhaust fan louvers.

Description:

The 2C DG room ventilation system consists of three roof fans for exhausting heat during operation and shutdown of the SBO diesel generator. The A and B fans exhaust heat from the room during the operating cycle, and the C fan exhaust heat from the room during the shutdown cycle. One of the fans (A or B) and half of the intake air wall louvers are capable of maintaining the room temperature below shutdown (104 °F) and operating design room temperature (122 °F) limits. The C fan starts when the temperature inside the room reaches 75 ºF, the A fan starts when the temperature inside the room reaches 80 ºF, and the B fan starts when the temperature inside the room reaches 105 °F.

During a walkdown of the 2C DG room, the team noted that the A and C fan were running, the B fan was rotating backwards, and the louvers for the B fan were stuck partially open. The licensee followed up on the teams observation and noted that the louvers for the A fan were also stuck partially open when it was not running. This configuration would allow the running fan to short-cycle air through the open louver.

Additionally, the non-running fan that was rotating backwards may not auto-start due to the tripping of thermal overloads. The licensee performed an operability/functionality assessment, and determined that with the A and C fan running, and the B fan spinning backwards, the 2C DG remained capable of performing its blackout function with reduced ventilation margin. The licensees compensatory measures placed the control switch for the A and B fans in manual until repairs are performed. The licensee entered this issue into their CAP as CRs 351580, CR 349883, and CR 355130.

The team reviewed the preventive maintenance and corrective action history for the 2C DG room exhaust louvers and noted that the last annual PM performed on the exhaust louvers was completed in 2010. The team also noted that the work instructions for the annual PM included steps to clean and lubricate the exhaust fan louvers, but did not include steps to inspect the condition of the louvers for indications of degradation or otherwise assess their functional capability. The PM completed in 2010 did not identify evidence of louver degradation. However, the team noted that previous annual inspections of the same louvers resulted in CRs written in 2008 and 2009 to address deficient conditions, but no repairs had been made to correct those issues. Based on these observations, and the degraded condition of the louvers, the team concluded that the PM program for the 2C DG exhaust fan louvers had been ineffective in providing assurance that the components would remain capable of performing their intended function.

Analysis:

The failure to perform condition monitoring or otherwise implement an appropriate preventive maintenance program for the 2C DG A and B exhaust fan louvers was a performance deficiency. This performance deficiency was more than minor because it was associated with equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform condition monitoring or otherwise implement an appropriate preventive maintenance program for the 2C DG A and B room exhaust fan louvers challenged the assurance that these components would remain capable of performing their intended functions. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 SDP screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. Because the licensee had initiated CRs in 2008 and 2009 for the 2C DG room exhaust louvers, and repairs were not made in a timely manner to address the issue, this finding was assigned a cross-cutting aspect in the corrective action program component of the problem identification and resolution area P.1(d).

Enforcement:

10 CFR 50.65(a)(1) states, in part, that licensees shall monitor the performance or condition of structures, systems and components (SSCs) within the scope of the rule as defined by 10 CFR 50.65(b), against license established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended function.

10 CFR 50.65(a)(2) states, in part, that monitoring as specified in 10 CFR 50.65(a)(1) is not required where it has been demonstrated that the performance or condition of an SSC is being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remains capable of performing its intended function.

Contrary to the above, since September 2010, the licensee had failed to demonstrate that the performance or condition of the 2C DG A and B exhaust fan louvers had been effectively controlled through the performance of appropriate preventive maintenance and did not otherwise monitor performance against licensee established goals. Because the violation was of very low safety significance and it was entered into the licensees CAP, this violation is being treated as an NCV consistent with the NRC Enforcement Policy: NCV 05000348, 364/2011010-07, Failure to Monitor or Perform Effective Preventive Maintenance on the 2C EDG Exhaust Fan Louvers.

.2.7 Turbine Driven Auxiliary Feedwater (TDAFW) Pump - Q2N23P0002

a. Inspection Scope

The team reviewed the plants TS, UFSAR, FSD, and P&IDs to establish an overall understanding of the design bases of the TDAFW. Design calculations and test data were reviewed to verify that design basis capability, and flow rates had been appropriately translated into these documents. The team concentrated its efforts on the pumps capability of performing its safety function (i.e., delivering the required flow rate to the steam generators at the prescribed design pressure). Records of surveillance testing and maintenance activities, and applicable corrective actions were examined to verify that potential degradation or low margin design issues were being monitored, prevented and/or corrected. TDAFW walkdowns were conducted to verify that the installed configurations would support its design basis function under accident conditions and had been maintained to be consistent with design assumptions and to visually inspect the material condition of the pumps. Control panel indicators were observed and operating procedures reviewed to verify that TDAFW operation and alignments were consistent with design and licensing basis assumptions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. The team reviewed the licensees severe weather procedure to verify that adequate guidance exists for operators to isolate the AFW minimum flow recirculation line during a postulated severance of the line as described in design bases calculations. The team performed a walkdown of local manual actions associated with isolation of the AFW minimum flow recirculation to verify the feasibility of the directed actions.

b.

.1 Findings

Introduction:

A green NRC identified NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to establish an adequate test procedure used to demonstrate that the TDAFW pump discharge check valves were capable of performing their design basis function. The test procedure was inadequate in providing assurance that the AFW system was capable of providing the required design basis flowrates to the SGs with reverse flow into an idle TDAFW pump via the discharge check valves.

Description:

Unit 1 and Unit 2 procedures, FNP-1/2-STP-22.30, Auxiliary Feedwater Pump Discharge Check Valve Reverse Flow Closure Operability Test, Version 6.1 and 5, were used by the licensee to demonstrate the closure of the TDAFW pump discharge check valves Q1/2N23V002D, Q1/2N23V002F, Q1/2N23V002H, and Q1/2N23V003.

The function of the check valves is to prevent reverse flow into the TDAFW pump when the pump is idle. The test acceptance criteria for check valve back leakage were less than 5 gallons per minute (gpm). Excessive back leakage through the pump discharge check valves would lower AFW system flowrate to the SGs below the flows required in the safety analysis for design basis events. The team identified the following deficiencies in the test procedure:

The team noted that the Unit 1 calculation 40.02, Verification of AFW Flow Bases, Rev.

4, and the Unit 2 calculation 38.04, Verification Of AFW Flow Bases, Rev. 4 identified a main feed water line break accident with the failure of the TDAFW pump as the most limiting case in terms of design flow margin to the other two SGs. Calculation 40.02 identified a flow margin of 2.21 gpm per SG for SG A and B, and calculation 38.4 identified a flow margin of 0.82 gpm per SG for SG A and C. Based on this information, the team concluded that the total acceptable leakage through the TDAFW pump discharge check valves was 4.21 gpm for Unit 1 and 1.64 gpm for Unit 2.

The team also noted that the check valves are tested at demineralized water system pressure which is approximately 80 to 126 psig. The actual system operating pressure these check valves would see is approximately 1130 psia. The higher operational differential pressure across the check valves could reasonably result in back leakage of more than 5 gpm at the higher design pressures. The team concluded that the procedure for the TDAFW discharge check valves was inadequate in that the 5 gpm acceptance criteria was non conservative with respect to the safety analysis margins available and the acceptance criteria was non conservative with respect to system pressure. This issue was entered into the licensees CAP as CR 348795.

The team reviewed the most recently completed test results for the TDAFW pump discharge check valves which indicated no evidence of back leakage (0 gpm observed).

The team concluded that, at the time of the inspection, the function of the check valves was not degraded.

Analysis:

The failure to develop an adequate test procedure which demonstrated that TDAFW pump discharge check valves were capable of performing their design basis function was a performance deficiency. This performance deficiency was more than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the acceptance criteria used in the test procedure was non-conservative when compared to the flowrates required by the accident analyses, and the test procedure was performed at lower system pressures (which were not representative of actual design conditions) without modifying the acceptance criteria. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team used the mitigating systems column to perform a Phase 1 SDP screening and determined the finding to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. Because the test procedure did not contain complete, accurate, and up-to-date information consistent with the system design basis safety analysis, this finding is assigned a cross-cutting aspect in the resources component of the human performance area H.2(c).

Enforcement:

10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and Procedures shall include appropriate quantitative or qualitative acceptance criteria. Contrary to the above, since April 2010, the licensee

(1) failed to establish a procedure for testing the leakage of the TDAFW discharge check valves that was appropriate to the circumstances, and

(2) failed to use appropriate quantitative or qualitative acceptance criteria. Specifically, the procedure for testing the TDAFW discharge check valves was inadequate in that the acceptance criteria was non conservative with respect to the safety analysis margins available for flow and the acceptance criteria was non conservative with respect to system pressure. Because the violation was of very low safety significance and it was entered into the licensees CAP, this violation is being treated as an NCV consistent with the NRC Enforcement Policy:

NCV 05000348, 364/2011010-08, Failure to Develop an Adequate Procedure to Test the TDAFW Pump Discharge Check Valves.

b.

.2 Findings

Introduction:

The team identified an URI regarding the licensees identification and evaluation of corrective actions taken to address AFW pump suction check valve oscillations.

Description:

In 2005 the licensee retained services of Kalsi Engineering to perform an analysis of the effects of partially open TDAFW pump suction check valves (2446, Kalsi Engineering, Auxiliary Feedwater Check Valve Analysis for Farley Nuclear Plant, Rev.

0, January 24, 2006). The engineering analysis determined that for 100 gpm of flow the check valve opening was 0.7 degrees with a disc peak-to-peak oscillating angle of 1.89 degrees. For 400 gpm of flow the check valve opening was 10.4 degrees with a disc peak-to-peak oscillating angle of 7.58 degrees. The analysis concluded that for each flow condition oscillation would not result in significant hinge pin wear. The scope of this review was limited to the TDAFW pump suction check valves only and the MDAFW pump suction check valves were not included in the analysis. The team noted that the 6 MDAFW pump suction check valves are the same model as the 8 TDAFW pump suction check valves. The teams review of the operating conditions for the AFW pump suction check valves identified the following observations:

• The quarterly inservice testing of the MDAFW pumps is performed at or near minimum flow conditions. The flow was not measured during the test; only the pumps differential pressure was monitored. The AFW system functional description indicates that this flow was approximately 50 gpm. The team determined that at this MDAFW pump flowrate, the check valve flow velocity was approximately 0.56 ft/sec, which was less than the TDAFW approximate velocity of 0.64 ft/sec at 100 gpm.

However, since MDAFW pump flow was not monitored during testing, any pump flow instability due to suction check valve oscillations may not be revealed during the current method of testing.

• Degradation (partial sticking) of the MDAFW pump suction check valves may not be identified by routine testing. If these valves are partially stuck open, then functional testing by either the quarterly surveillances or the comprehensive surveillance test will not reveal this condition, since higher-than-design-basis CST levels would mask the valve being partially stuck open.

• On October 2, 2008, the licensee initiated CR 2008110018, which identified that while the TDAFW pump was running in a minimum flow alignment, the pump exhibited flow fluctuations at the flow rates below approximately 230 gpm. The licensee attributed this condition to a partially open TDAFW suction check valve (Q2N23V0006). During the subsequent troubleshooting, the licensee disassembled this check valve on October 28, 2008, and verified that the check valve was functioning properly without any abnormal wear indications. Proposed corrective actions considered replacement of the swing check valves with the in-line check valves. However, because of the reliability concerns and difficulty in performing maintenance inspections on that type of check valve, this modification was not pursued. Additionally, the team noted that the MDAFW pump suction check valves were not inspected.

Based on a review of the operating and corrective action history related to AFW check valve oscillations for this issue over the last 5 years, the team concluded that there was not sufficient evidence to conclude that the MDAFW pump suction check valves would not be subject to the same oscillation issue that was observed on the TDAFW pump.

Additionally, the team noted that the check valve oscillating condition that had been previously evaluated for the TDAFW pumps, had not been evaluated for MDAFW pumps.

Summary: The team determined that additional information and/or evaluation from the licensee regarding the current condition of the MDAFW pump suction check valves would be required to confirm that the check valves were not adversely impacting the AFW system design basis function. Additionally, the team concluded that this additional evaluation of the current condition of MDAFW pump suction check valves would be required to determine if this issue resulted in a more than minor performance deficiency.

(URI 05000348, 364/2011010-09, Evaluation of MDAFW Pump Suction Check Valves)

.2.8 MDAFW Pump (Electrical) - Q2N23M0001A

a. Inspection Scope

The team reviewed the plants FSAR, TS and FSD to establish an overall understanding of the design bases of the controls for the MDAFW pumps. Electrical drawings and site procedures were reviewed to verify that the design bases and design assumptions had been appropriately translated into these documents. Vendor documentation, system health reports, and problem history were reviewed in order to verify that the MDAFW pump controls were being properly maintained. The team reviewed the licensees alarm response procedures associated with AFW flow alarms to verify that setpoints were consistent with design bases documents. The team reviewed maintenance documentation to verify that the components were calibrated.

b. Findings

A violation was identified regarding AFW pump minimum flow alarm setpoints and is documented in section 1R21.2.4 of this report.

.2.9 Reactor Trip and Bypass Breakers - Q1C11E0004ART(BBY)

a. Inspection Scope

The team reviewed the plants FSAR, TS and FSD to establish an overall understanding of the design bases of the reactor trip circuit breakers operation and actuation. Electrical drawings and site procedures were reviewed to verify that the design bases and design assumptions had been appropriately translated into these documents. Test procedures and results of previous testing and refurbishment activities were reviewed against FSDs to verify that acceptance criteria for tested parameters were supported by the accident analyses. Vendor documentation, system health reports, and maintenance history were reviewed in order to verify that the trip circuit breakers were being properly maintained.

Since the reactor trip and bypass circuit breakers were the same model (Westinghouse Model DS-416) that had been the subject of an earlier NRC Information Notice, the team also reviewed the actions that had been taken by the licensee in response to Information Notice (IN) 1992-29, Potential Breaker Mis-coordination Caused by Instantaneous Trip Circuitry.

b. Findings

No findings were identified.

.2.10 F 600 Volt Load Center - Q1R16B0008-AB

a. Inspection Scope

The team reviewed the plants FSAR, TS and FSD to establish an overall understanding of the design bases for the 1F 600 Volt Load Center. The team reviewed electrical drawings and site procedures to verify that the design bases and had been appropriately translated into these documents. The team reviewed the operation of the load center 1F key interlocks that were provided between the 4160 Volt supply breakers, the associated disconnects, and the 600 Volt feeder circuit breakers to verify that adequate guidance was provided to ensure correct alignment to one of the two 4160 Volt engineered safeguard busses 1F or 1G. The team also reviewed involved plant procedures to ensure that adequate guidance was provided for connecting power from load center 1F to one of the other 600 Volt load centers.

b. Findings

No findings were identified.

.2.11 1-2A Emergency Diesel Generator Voltage Regulator - QSR43A0501GENRG

a. Inspection Scope

The team reviewed the plants FSAR, TS and FSD to establish an overall understanding of the design bases of the emergency diesel generator (EDG) voltage regulator system.

Electrical drawings and site procedures were reviewed to verify that the design bases and design assumptions had been appropriately translated into these documents. The team reviewed system modifications over the life of the component to verify that the subject modifications did not degrade the components performance capability and were appropriately incorporated into relevant drawings and procedures. Component walk downs of the 1-2A EDG were conducted to verify that the installed configurations would support their design bases function under accident/event conditions and had been maintained to be consistent with design assumptions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing bases assumptions. Test procedures and results were reviewed against FSDs to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and that component replacement was consistent with equipment qualification life.

b. Findings

No findings were identified.

.2.12 Auxiliary Building Safety-Related Batteries - Q2R42E002A-A(B-B)

a. Inspection Scope

The team reviewed battery sizing and loading calculations to verify that loads do not exceed battery bank capacity. The team verified that the load profile bounded all accident scenarios. Also, the team reviewed short circuit calculations to verify that the duty cycle does not exceed the equipment protection ratings. The team reviewed performance tests to verify that the minimum voltage at the end of the test is the minimum voltage required by the most limiting component that has to actuate. In addition, a review of the service test was performed to verify that for the required current, the battery can provide the adequate voltage during an accident. The team reviewed equalizing procedures for the batteries to verify proper voltage. Selective one-line and schematic diagrams were reviewed to verify proper configuration of the 125 Volt Direct Current (VDC) electrical distribution system. The team performed a walkdown to verify material condition of the batteries and reviewed a sample of condition reports to confirm that the licensee adequately identifies, evaluates, and dispositions adverse conditions.

b. Findings

No findings were identified.

.2.13 Auxiliary Building Safety-Related Battery Chargers - Q2R42E0001A(B/C)

a. Inspection Scope

The team reviewed battery charger sizing calculations to verify that the chargers are capable of carrying the continuous load during a Design Basis Accident (DBA) and will charge the batteries to full capacity within required time. Also, the team reviewed the last two tests of the battery chargers to look for signs of degradation due to aging. A review of the ac voltage calculation was performed to assure satisfactory voltage to the chargers under worst-case conditions. In addition, the team verified that the ampere-hours returned to the battery were greater than the ampere hours removed plus the charging losses. The team performed a walkdown to verify material condition of the components and reviewed a sample of condition reports to confirm that the licensee adequately identifies, evaluates, and dispositions adverse conditions.

b. Findings

No findings were identified.

.2.14 SG Narrow Range (NR) Level Instrumentation - Q1C22LT0474-476(484-486/494-496)

a. Inspection Scope

The team reviewed instrument setpoint and uncertainty calculations, as well as calibration procedures and calibration test records to verify that the SG NR level instruments were in accordance with design bases documents. The last two completed calibration test records were reviewed to confirm that instrument setpoints were consistent with setpoint calculations. Also, the team reviewed a sample of condition reports to confirm that the licensee adequately identified and corrected adverse conditions. In addition, the team reviewed the maintenance history to verify actions were taken to correct and prevent problems.

b. Findings

No findings were identified.

.2.15 Reactor Coolant Pump Thermal Barrier Isolation Valves - Q2P17HV3184/3045

a. Inspection Scope

The team reviewed the isolation valves to verify their capability to perform the required design function. The review included the licensing and design basis of the valves, review of recent corrective actions, review of recent test procedures and test results, walkdowns of the valves and related instruments, and interviews conducted with responsible engineering personnel. The team reviewed the test procedures associated with the valves to verify the valves and instruments were being tested in accordance with the design bases. In addition, the team reviewed the maintenance history to verify actions were taken to correct and prevent problems. The team also conducted walkdowns of the valves and associated equipment to verify the material condition of the components.

b. Findings

No findings were identified.

.3 Operating Experience (4 Samples)

a. Inspection Scope

The team reviewed four operating experience issues for applicability at Farley Nuclear Plant. The team performed an independent review for these issues and where applicable, assessed the licensees evaluation and dispositioning of each item. The issues that received a detailed review by the team included:

• Generic Letter 1996-05, Periodic Verification of Design Basis Capability of Safety-Related Motor-Operated Valves

• NRC Information Notice 1992-29, Potential Breaker Miscoordination Caused by Instantaneous Trip Circuitry

• Generic Letter 2007-01, Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant Transients

• Bulletin 88-04, Potential Safety Related Pump Loss

b. Findings

No findings were identified.

4OA5 Other Activities

.1 Review of Degraded Voltage Protection Design and Licensing Bases

a. Inspection Scope

During the inspection period, the team reviewed the licensees degraded voltage protection design and licensing bases. The team reviewed functional system descriptions, technical specifications, corrective action program documents, licensee self-assessments, and safety evaluation reports related to degraded voltage protection.

The team evaluated the current degraded voltage design and licensing bases using the regulatory requirements specified in 10 CFR 50.55a(h)(2) and 10 CFR 50, Appendix A, General Design Criterion 17, Electric Power Systems. Additionally, the team evaluated the degraded voltage protection design using the staff positions provided in Standard Review Plan, NUREG-0800, (July 1981), and Branch Technical Positions (BTPs) of Appendix 8-A (PSB), containing BTP PSB-1, Adequacy of Station Electric Distribution System Voltages.

b. Findings and Observations

Introduction:

The team indentified an Unresolved Item (URI) regarding the licensees use of administrative controls in lieu of automatic degraded voltage protection to assure adequate voltage to safety-related equipment during design basis events.

Description:

The team noted that the degraded voltage protection system at Farley uses administrative controls to assure adequate voltage to safety-related equipment during design basis events. Farleys current system configuration, which relies on administrative actions, was recognized as a deviation from the guidance on degraded voltage protection provided in a NRC letter (dated June 2, 1977), but was accepted by the NRC in a Safety Evaluation Report (SER) (dated November 21, 1995). The licensee entered this issue into their corrective action program as CR 2011106624 on May 17, 2011.

This same issue is currently being assessed at plant Hatch, another Southern Company licensee, where the agency issued a backfit letter (Hatch Inspection Report 05000321/2011009 and 05000366/2011009, dated May 25, 2011). In the backfit letter, the staff concluded that the NRC was in error in accepting the use of administrative controls.

Summary: Because of the similarities of this issue for plants Farley and Hatch, this issue is unresolved pending completion of the appeal process that is afforded to Southern Company for plant Hatch. (URI 05000348, 364/2011010-10, Administrative Controls in lieu of Automatic Actions for Degraded Voltage Protection.

4OA6 Meetings, Including Exit

On September 30, 2011, the team discussed the status of the inspection with Mr. Tom Lynch and other members of the licensees staff. On December 8, 2011, the team presented the inspection results to Mr. Todd Youngblood and other members of the licensees staff. Proprietary information that was reviewed during the inspection was returned to the licensee or destroyed in accordance with prescribed controls.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

B. Oldfield, Licensing

B. Nobles, Site Design

M. Byrd, Design Engineering Supervisor

NRC personnel

R. Nease, Chief, Engineering Branch Chief 1, Division of Reactor Safety, RII

E. Crowe, Senior Resident Inspector, Division of Reactor Projects, Farley Resident Office

LIST OF ITEMS

Opened and Closed

05000348, 364/2011010-01 NCV Failure to Implement Design Control Measures to Verify the Adequacy of CST Design (Section 1R21.2.3)

05000348, 364/2011010-02 NCV Failure to Implement Design Control Measures to Verify the Adequacy of AFW Design (Section 1R21.2.3)

05000348, 364/2011010-03 NCV Failure to Provide Adequate Procedural Guidance for Controlling Steam Generator and Pressurizer Level During Loss of Air Events (Section 1R21.2.3)

05000348, 364/2011010-06 NCV Failure to Correctly Translate the Design Basis into Procedures for Minimum CCW Flow to the RHR Seal Coolers and Minimum Flow Requirements for the AFW Pumps (Section 1R21.2.4)

05000348, 364/2011010-07 NCV Failure to Monitor or Perform Effective Preventive Maintenance on the 2C EDG Exhaust Fan Louvers (Section 1R21.2.6)

05000348, 364/2011010-08 NCV Failure to Develop an Adequate Procedure to Test the Turbine Driven Auxiliary Feedwater Pump Discharge Check Valves (Section 1R21.2.7)

Opened

05000348, 364/2011010-04 URI Evaluation of CST Vortex Effect on AFW Pump Minimum Submergence (Section 1R21.2.3)

05000348, 364/2011010-05 URI Non-Conservative Assumptions Regarding AFW Net Positive Suction Head (Section 1R21.2.3)

05000348, 364/2011010-09 URI Evaluation of MDAFW Pump Suction Check Valves (Section 1R21.2.7)

05000348, 364/2011010-10 URI Administrative Controls in lieu of Automatic Actions for Degraded Voltage Protection (Section 4OA5.1)