ML11291A094

IR 05000266-11-009, 05000301-11-009; on 08/01/11 - 9/2/11, Point Beach Nuclear Plant, Units 1 and 2; Component Design Bases Inspection (CDBI)

ML11291A094
Person / Time
Site:	Point Beach
Issue date:	10/17/2011
From:	Stone A M NRC/RGN-III/DRS/EB2
To:	Meyer L Point Beach
References
IR-11-009
Download: ML11291A094 (38)

See also: IR 05000266/2011009

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 2443 WARRENVILLE ROAD, SUITE 210 LISLE, IL 60532-4352 October 17, 2011 Mr. Larry Meyer Site Vice President NextEra Energy Point Beach, LLC

6610 Nuclear Road Two Rivers, WI 54241 SUBJECT: POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2; COMPONENT DESIGN BASES INSPECTION (CDBI) REPORT 05000266/2011009; 05000301/2011009 Dear Mr. Meyer: On September 2, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed a Component Design Bases Inspection (CDBI) at your Point Beach Nuclear Plant. The enclosed report documents the results of this inspection, which were discussed on September 2, 2011, with Mr. T. Vehec and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.

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

personnel. Based on the results of this inspection, four NRC-identified findings of very low safety significance were identified. Three of the findings involved violations of NRC requirements.

However, because of their very low safety significance, and because the issues were entered into your corrective action program, the NRC is treating the issues as Non-Cited Violations (NCVs) in accordance with Section 2.3.2 of the NRC Enforcement Policy. If you contest the subject or severity of this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear

Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a

copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III,

2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement,

U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Point Beach Nuclear Plant. In addition, 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 III, and the NRC Resident Inspector at the Point Beach Nuclear Plant.

L. Meyer -2- 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 System (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). Sincerely,

/RA/

Ann Marie Stone, Chief Engineering Branch 2

Division of Reactor Safety Docket Nos. 50-266; 50-301 License No. DPR-24; DPR-27 Enclosure: Inspection Report 05000266/2011009; 05000301/2011009 w/Attachment: Supplemental Information cc w/encl: Distribution via ListServ

Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION III Docket No: 05000266; 05000301 License No: DPR-24; DPR-27 Report No: 05000266/2011009; 05000301/2011009

Licensee: NextEra Energy Point Beach, LLC

Facility: Point Beach Nuclear Plant, Units 1 and 2

Location: Two Rivers, WI

Dates: August 1 through September 2, 2011 Inspectors: Alan Dahbur, Senior Engineering Inspector, Lead Caroline Tilton, Senior Engineering Inspector, Mechanical

Mohammad Munir, Engineering Inspector, Electrical

Carl Moore, Operations Inspector

John Bozga, Civil Structural Inspector Jerry Nicely, Electrical Contractor Bill Sherbin, Mechanical Contractor Trainee: Cimberly Nickell, Nuclear Safety Professional Development Program, NRR Approved by: Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety

1 Enclosure SUMMARY OF FINDINGS IR 05000266/2011009, 05000301/2011009; 8/01/2011 - 9/02/2011; Point Beach Nuclear Plant, Units 1 and 2; Component Design Bases Inspection (CDBI). The inspection was a 3-week onsite baseline inspection that focused on the design of components. The inspection was conducted by regional engineering inspectors and two

consultants. Four Green findings were identified by the inspectors. Three of the findings were

considered Non-Cited Violations (NCVs) of NRC regulations. The significance of most findings

is indicated by their color (Green, White, Yellow, Red) using 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," Revision 4, dated December 2006. A. NRC-Identified and Self-Revealed Findings Cornerstone: Initiating Events * Green. The inspectors identified a finding of very low safety significance involving the licensee's failure to meet the requirements of the American Institute of Steel Construction (AISC) Specification. Specifically, the licensee's design basis calculation failed to ensure the turbine building structural steel floor beams met the AISC

specification. This finding was entered into the licensee's corrective action program. No

violation of NRC requirements was identified. The performance deficiency was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective to limit the likelihood of those events that

upset the plant's stability and challenged critical safety functions during shutdown, as

well as power operations. The finding screened as very low safety significance (Green), because the transient initiator would not contribute to both the likelihood of a reactor trip

and the likelihood that mitigation equipment or functions will not be available. This finding had a cross-cutting aspect in human performance and work practice because the licensee did not ensure effective supervisory and management oversight of work

activities, including contractors, such that nuclear safety was supported. Specifically, the

licensee failed to have adequate oversight of design calculation and documentation for establishing structural adequacy of the turbine building structural steel beams at EL. 44'-0." H.2(c) (Section 4OA5.1.b.(2)) Cornerstone: Mitigating Systems * Green. The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," involving the licensee's failure to correctly translate design basis assumptions

into procedures or instructions. Specifically, the licensee failed to monitor average

outside air temperature which was one of the design input criteria for the temperature

heat-up calculation associated with rooms which housed safety-related equipment. This finding was entered into the licensee's corrective action program.

2 Enclosure The performance deficiency was associated with Mitigating System Cornerstone and determined to be more than minor because, if left uncorrected, it could lead to a more

significant safety concern. The finding screened as very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting

aspect in the area of human performance, resources because the licensee did not

ensure adequate training and qualification of personnel. Specifically, the licensee failed to adequately train licensed operators to ensure adequate knowledge with respect to the interface between functionality of a non-safety system component and the impact of a failure on the operability of safety-related equipment. H.2(b). (Section 1R21.3.b.(1)) * Green. The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to ensure a minimum AFW flow of 275 gpm as specified in the

accident analysis for the Loss of Normal Feedwater event. This finding was entered into the licensee's corrective action program. The performance deficiency was associated with the Mitigating Systems Cornerstone attribute of design control and was determined to be more than minor because, if left

uncorrected, it would have the potential to lead to a more significant safety concern.

Specifically, an AFW flow rate of less than 275 gpm as specified in the procedures did not ensure the pressurizer would not become water solid and cause an over-pressure condition within the Reactor Coolant System during the Loss of Normal Feedwater. The

finding screened as of very low safety significance (Green) because the finding was not

a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of human performance, resources because the licensee did not maintain design documentation in

a complete and accurate manner. Specifically, the licensee failed to maintain

Emergency Procedures consistent with the design basis analysis for LONF. H.2(c). (Section 1R21.6.b.(1)) Cornerstone: Barrier Integrity * Green. The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," involving the licensee's failure to ensure the Containment Spray Pipe Support

2S-249 and Containment Spray Pipe Anchor 2A-35 meet Seismic Category I requirements. This finding was entered into the licensee's corrective action program. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect

the public from radionuclide releases caused by accidents or events. This finding is of

very low safety significance (Green) because there was no actual barrier degradation.

The inspectors did not identify a cross-cutting aspect associated with this finding because this was a legacy design issue; and therefore, was not reflective of current performance. P.1(a). (Section 4OA5.1.b.(1))

3 Enclosure B. Licensee-Identified Violations Violations of very low safety significance that were identified by the licensee have been reviewed by inspectors. Corrective actions planned or taken by the licensee have been entered into the licensee's corrective action program. These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.

4 Enclosure REPORT DETAILS 1. REACTOR SAFETY Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity 1R21 Component Design Bases Inspection (71111.21) .1 Introduction The objective of the component design bases inspection is to verify the design bases have been correctly implemented for the selected risk significant components and that operating procedures and operator actions are consistent with design and licensing

bases. As plants age, their design bases may be difficult to determine and an important design feature may be altered or disabled during a modification. The Probabilistic Risk-Assessment (PRA) model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems, and Barrier Integrity

cornerstones for which there are no indicators to measure performance. Specific documents reviewed during the inspection are listed in the Attachment to the report. .2 Inspection Sample Selection Process Point Beach Nuclear Plant implemented major modifications to the existing Auxiliary Feedwater System in support of the extended power uprate and to resolve other system low margin issues. The modification included the addition of two higher capacity motor driven pumps and their associated valves and piping. The inspectors used information contained in the licensee's PRA, the Point Beach's Standardized Plant Analysis Risk

Model as the basis for component selection from the AFW System. Using the system approach as specified in the inspection procedures, a number of risk significant

components were selected for the inspection including components used to support the AFW system. The inspectors also used additional component information such as a margin assessment in the selection process. This design margin assessment considered

original design reductions caused by design modification, power uprates, or reductions

due to degraded material condition. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as performance test results, significant corrective actions, repeated maintenance activities,

Maintenance Rule (a)(1) status, components requiring an operability evaluation, NRC

resident inspector input of problem areas/equipment, and system health reports. Consideration was also given to the uniqueness and complexity of the design, operating

experience, and the available defense in depth margins. A summary of the reviews performed and the specific inspection findings identified are included in the following sections of the report.

5 Enclosure The inspectors also identified procedures and modifications for review that were associated with the selected components. In addition, the inspectors selected operating experience issues associated with the selected components. This inspection constituted 22 samples as defined in IP 71111.21-05. .3 Component Design a. Inspection Scope The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), Technical Specifications (TS), design basis documents, drawings, calculations and other available design basis information, to determine the performance requirements of the selected components. The inspectors used applicable industry standards, such as the American

Society of Mechanical Engineers (ASME) Code, Institute of Electrical and Electronics

Engineers Standards and the National Electric Code, to evaluate acceptability of the systems' design. The NRC also evaluated licensee actions, if any, taken in response to NRC issued operating experience, such as Bulletins, Generic Letters (GLs), Regulatory Issue Summaries (RISs), and Information Notices (INs). The review was to verify the

selected components would function as designed when required and support proper

operation of the associated systems. The attributes that were needed for a component

to perform its required function included process medium, energy sources, control systems, operator actions, and heat removal. The attributes to verify the component condition and tested capability was consistent with the design bases and was

appropriate may include installed configuration, system operation, detailed design,

system testing, equipment and environmental qualification, equipment protection, component inputs and outputs, operating experience, and component degradation. For each of the components selected, the inspectors reviewed the maintenance history, preventive maintenance activities, system health reports, operating experience-related information, vendor manuals, electrical and mechanical drawings, and licensee corrective action program documents. Field walkdowns were conducted for all

accessible components to assess material condition and to verify the as-built condition was consistent with the design. Other attributes reviewed are included as part of the scope for each individual component. The following 18 components were reviewed: * 4.16 kV Switchgear Bus (2A06): The inspectors reviewed electrical distribution system load flow/voltage drop, degraded voltage protection, short-circuit, and electrical protection and coordination associated with the safety-related 4.16 KV

Bus. This review was conducted to assess the adequacy and appropriateness of

design assumptions, and to verify the bus capacity was not exceeded and bus voltages remained above minimum acceptable values under design basis conditions. The review included switchgear's protective device settings and

breaker ratings to ensure the selective coordination was adequate for protection

of connected equipment during worst-case, short-circuit conditions. The 125Vdc

voltage calculations were reviewed to determine if adequate voltage would be available for the breaker open/close coils and spring charging motors during

6 Enclosure events. The station's interface and coordination with the transmission system operator for plant voltage requirements and notification set points were reviewed.

The inspectors evaluated selected portions of the licensee's response to NRC Generic Letter (GL) 2006-02, "Grid Reliability and the Impact on Plant Risk and the Operability of Offsite Power," dated February 1, 2006. The inspectors

reviewed the degraded and loss of voltage relay protection schemes and bus

transfer schemes between offsite power supplies and the associated emergency

diesel generators. In addition, the inspectors reviewed the preventive

maintenance inspection and testing procedures to verify the breakers were maintained in accordance with industry and vendor recommendations. System health reports, component maintenance history, and licensee's corrective action

program reports were reviewed to verify correction of potential degradation and

deficiencies were appropriately identified and resolved. The inspectors reviewed

selected industry operating experiences and plant actions to address the applicable issues to ensure the appropriate insights from operating experience have been applied. * 480 VAC Switchgear Bus (2B-04): The inspectors inspected the 480V switchgear to verify it would operate during design basis events. The inspectors reviewed selected calculations for electrical distribution system load flow/voltage drop, short-circuit, and electrical protection and coordination. The adequacy and

appropriateness of design assumptions and calculations were reviewed to verify the bus and circuit breaker capacity was not exceeded and bus voltages

remained above minimum acceptable values under design basis conditions. The switchgear's protective device settings and breaker ratings were reviewed to ensure the selective coordination was adequate for protection of connected

equipment during worst-case short-circuit conditions. To ensure the breakers

were maintained in accordance with industry and vendor recommendations, the

inspectors reviewed the vendor manuals, preventive maintenance inspection, and testing procedures. The 125Vdc voltage calculations were reviewed to determine if adequate voltage would be available for the breaker open/close

coils during events. System health reports, component maintenance history

and licensee's corrective action program reports were reviewed to verify

correction of potential degradation and deficiencies were appropriately identified

and resolved. The inspectors reviewed selected industry OE and any plant actions to address the applicable issues to ensure the appropriate insights from operating experience have been applied. Finally, the inspectors performed a

visual non-intrusive inspection of observable portions of the safety-related 480V

Switchgear Bus 2B-04 to assess the installation configuration, material condition, and the potential vulnerability to hazards. * 480 VAC Motor Control Center (MCC 2B-42): The inspectors inspected the 480V MCC to verify it would operate during design basis events. The inspectors reviewed selected calculations for electrical distribution system load flow/voltage drop, short-circuit, and electrical protection and coordination. The adequacy and appropriateness of design assumptions and calculations were reviewed to verify the bus and circuit breaker capacity was not exceeded and bus voltages

remained above minimum acceptable values under design basis conditions. The

7 Enclosure MCC's protective device settings and breaker ratings were reviewed to ensure the selective coordination was adequate for protection of connected equipment

during worst-case short-circuit conditions. To ensure the breakers were maintained in accordance with industry and vendor recommendations, the inspectors reviewed the vendor manuals, preventive maintenance inspection,

and testing procedures. System health reports, component maintenance history

and licensee's corrective action program reports were reviewed to verify

correction of potential degradation and deficiencies were appropriately identified

and resolved. The inspectors reviewed selected industry OE and any plant actions to address the applicable issues to ensure appropriate insights from operating experience have been applied. Finally, the inspectors performed a

visual non-intrusive inspection of observable portions of the safety-related 480V

MCC 2B-42 to assess the installation configuration, material condition, and the potential vulnerability to hazards. * 125 VDC Battery (D06): The inspectors reviewed various electrical calculations and analyses associated with the safety-related battery to verify the battery was designed and capable to perform its function and provide adequate voltage for

required loads during design basis accident and station blackout (SBO) event. These calculations included battery sizing and capacity, voltage drop, minimum voltage, hydrogen generation, SBO loading, and battery room transient

temperature. The inspectors also reviewed a sampling of completed weekly,

monthly, semi-annual surveillance tests including performance discharge tests,

and modified performance tests. The review was performed to ascertain that acceptance criteria were met and performance degradation would be identified. * 125 VDC Bus (D02): The inspectors reviewed various electrical calculations and analysis associated with the safety-related 125 Vdc bus including voltage drop, short circuit and fuse interrupting ratings to verify sufficient power and voltage

was available at the safety-related equipment supplied by this bus to perform

their safety function; and the interrupting ratings of the fuses were well above the

calculated short circuit currents. The inspectors also reviewed schematic and elementary diagrams for motor control logic to ensure adequate voltage would be available for the control circuit components under all design basis conditions. * 1/2P-53 Pumps Main Feeder Breakers (1A52-83 and 2A52-68): The inspectors inspected the 4kV circuit breakers 1P-53 and 2P-53 to verify the capability to meet the design basis requirements, which is to supply power to the safety-

related motor driven Auxiliary Feedwater Pump motors (MDAFWP) 1P-53 and

2P-53, MDAFWP 1P-53 is fed from 4160V Safeguards Bus Train B 1A-06

through 4kV breaker 1A52-83. MDAFWP 2P-53 is fed from 4160V Safeguards Bus Train A 2A-05 through 4kV breaker 2A52-68. The inspectors reviewed one line diagrams and vendor equipment data to confirm the breaker ratings were

sufficient to meet design basis conditions. The inspectors reviewed the electrical

analyses for loading and protection and coordination requirements to confirm the

adequacy of the protective device settings for motor operation and circuit protection and coordination with upstream power supplies. The inspectors reviewed manufacturer vendor manuals, periodic maintenance and testing

8 Enclosure practices to ensure the equipment is maintained in accordance with industry practices. The associated breaker closure and opening control logic diagrams

and the 125Vdc voltage calculations were reviewed to verify adequate voltage would be available for the breaker open/close coils and spring charging motors under accident/event conditions. System health reports, component

maintenance history and licensee's corrective action program reports were

reviewed to verify correction of potential degradation and deficiencies were

appropriately identified and resolved. The inspectors reviewed selected industry

OE and any plant actions to address the applicable issues to ensure appropriate insights from operating experience have been applied. The inspectors performed a visual non-intrusive inspection of 4kV circuit breakers 1P-53 and 2P-53 to

assess the installation configuration, material condition, and potential

vulnerability to hazards. * Motor-Driven AFW Pump (2P-53): The inspectors reviewed design documents, including drawings and calculations to determine the design requirements for the new MDAFW pump. The inspectors reviewed the Safety Analysis Report, and recent addendum, to determine the licensing basis requirements for the system, in order to determine the hydraulic requirements for the pump. Hydraulic analyses were reviewed to verify adequacy of Net Positive Suction Head (NPSH)

and to verify the adequacy of surveillance test acceptance criteria for pump

minimum discharge pressure at required flow rate. The results of the inservice testing (IST) performed during start-up of 2P-53, were reviewed to verify acceptance criteria were met and performance degradation would be identified.

Pump actuation logic test results were reviewed to ensure the MDAFW pump

would start in accidents and events as described in the UFSAR. The inspectors

reviewed condensate storage tank (CST) design criteria, including usable volume

calculations to ensure the MDAFW pump, in conjunction with the turbine driven AFW pump had adequate water supply to prevent vortexing prior to switchover of pump suction to the service water supply. Seismic calculation of the pump

mounting bolts was reviewed for adequacy. Condition Reports were reviewed to

ensure problems were identified and corrected in a timely manner. The

inspectors reviewed the pipe stress analysis and pipe support calculations associated with these pumps to verify the pumps meet the design basis requirements. * 2P-53 Pump Minimum Flow Valves (2AF-04073A/B): The MDAFW pump has two minimum flow control valves (in parallel). Minimum pump flow is required to remove pump heat, and ensure hydraulic stability when the pump is running.

This review included design analyses of the valves and associated air receiver tank to verify the capability of the valves to perform their required function. Specifically, the inspectors reviewed air-operated valve thrust calculations, reviewed the required air pressure to open the valve, and reviewed the capacity

and allowable leakage limits of the associated air receiver to verify the capability

of the valves to perform their function when required. The inspectors verified the

valves were sized to provide adequate pump minimum flow to preclude pump

degradation and heat-up when operating under minimum flow conditions. The

9 Enclosure inspectors reviewed start-up testing of the P-53 pumps to ensure the minimum flow valves were functionally tested to open and close at the required setpoints. * 2P-53 Pump Discharge Flow Control Valves (2AF-04074A/B): These valves have an automatic function to throttle MDAFW pump discharge flow to each steam generator to maintain a set discharge flow rate. This review included

design analyses of the valves and associated air receiver tank to verify the

capability of the valves to perform their required function. Specifically, the inspectors reviewed air-operated valve thrust calculations, reviewed the required air pressure to open the valve, and reviewed the capacity and allowable leakage

limits of the associated air receiver to verify the capability of the valves to perform

their function when required. The inspectors reviewed start-up testing of the 2P-

53 pump to ensure the discharge flow control valves were functionally tested to throttle flow to the steam generators. The inspectors also reviewed the design of the valve internals to ensure potential blockage by debris would not inhibit AFW flow to the steam generators. * Service Water Cross-Tie Valve to 2P-53 Pump Suction Line (2AF-4067): The inspectors reviewed the service water cross-tie valve to verify it was capable of performing its design basis requirement of providing safety grade water to the

MDAFW pump suction line when required. The review included service water

hydraulic calculations and MOV analysis to ensure thrust and torque limits and

actuator settings were appropriate. The inspectors reviewed start-up testing of the 2P-53 pump to ensure the valve was functionally tested to stroke open based on minimum CST level, and pump low suction pressure instrumentation.

Additionally, the inspectors reviewed the MOV voltage drop calculation to ensure

appropriate voltage values were used in the thrust calculation. The inspectors

also reviewed surveillance procedures, and results of the periodic flushing of

service water suction lines to the valve to ensure the lines are maintained free of debris. In addition, the inspectors reviewed electrical calculation to verify the adequacy of feeder circuit including breaker, cable, breaker settings, electrical

schematic, control switch settings, 125 VDC power and control voltage drop,

thermal overload relay settings, thermal overload relay testing, breaker/fuse coordination. * Turbine Driven Auxiliary Feedwater (TDAFW) Pump/Turbine (2P-29): The inspectors reviewed the AFW system to verify the pump and associated peripherals could meet the design and performance requirements identified in the

AFW system design/licensee's basis and the FSAR. The inspection included a review of required flows for transients and postulated SBO events, as well as minimum flow provisions. The inspectors evaluated flow calculations, net

positive suction head (NPSH) calculations, and test data to ensure the design

basis requirements were met. The inspectors reviewed completed surveillance

test results to verify the acceptance criteria and test results demonstrated pump operability was being maintained. The inspectors also reviewed room heat-up calculations, procedures used to mitigate the effects of loss of normal ventilation,

and surveillances conducted on temporary fan units. In addition, the inspectors

10 Enclosure reviewed normal and abnormal operating procedures to ensure these would perform their objectives. * TDAFW 2P-29 Minimum Flow Valve (2AF-4002): The inspectors reviewed information related to the air-operated valve (AOV) installed in the minimum flow line of the TDAFW pump. This review included inservice test procedures and

results to verify the capability of the valve to perform its required function under

postulated accident conditions. The inspectors also reviewed the design of the instrument air supply line and accumulator to verify the valve would function as designed. * Suction Pressure Transmitters to AFW pumps (2PT-4044 and 2PT-4071): The inspectors reviewed the piping and instrumentation diagram (P&ID), Technical Specification requirements, setpoint calculation including the verification of

instrument and loop uncertainty, completed calibration procedures to ensure the transmitter was capable of functioning under design conditions. * Service Water Supply to TDAFW Pump 2P-29 (2AF-4006): The inspectors reviewed MOV calculations and analysis to ensure the valve was capable of functioning under design conditions. These included calculations for required thrust. Diagnostic testing and IST surveillance results, including stroke time,

were reviewed to verify acceptance criteria were met and performance

degradation could be identified. In addition, the inspectors reviewed electrical

calculation to verify the adequacy of feeder circuit including breaker, cable, breaker settings, electrical schematic, control switch settings, 125 VDC power and control voltage drop, thermal overload relay settings, thermal overload relay

testing, and breaker/fuse coordination. * TDAFW 2P-29 Bearing Oil Cooling (2MS-2090S): The inspectors reviewed information related to the bearing oil cooler on the turbine side of the TDAFW pump. The review included design configuration and specification. The

inspectors also evaluated the adequacy of the station's GL 89-13 program in

maintaining the heat removal efficiency of the bearing oil cooler. The inspectors reviewed a sample of completed surveillances to verify acceptance criteria were met and performance degradation could be identified. * TDAFW Pump 2P-29 Steam Supply Valves (2MS-2019 and 2MS-2020): The inspectors reviewed motor-operated valve (MOV) calculations and analysis to ensure the valves were capable of functioning under design conditions.

Diagnostic testing and IST surveillance results, including stroke time and

available thrust, were reviewed to verify acceptance criteria were met and performance degradation could be identified. * TDAFW Pump 2P-29 Discharge Valves (2AF-4000 and 2AF-4001): The inspectors reviewed motor-operated valve (MOV) calculations and analysis to ensure the valves were capable of functioning under design conditions. These

included calculations for required thrust and maximum differential pressure.

Diagnostic testing and IST surveillance results, including stroke time and

11 Enclosure available thrust, were reviewed to verify acceptance criteria were met and performance degradation could be identified. In addition, the inspectors

reviewed electrical calculation to verify the adequacy of feeder circuit including breaker, cable, breaker settings, electrical schematic, control switch settings, 125 VDC power and control voltage drop, thermal overload relay settings,

thermal overload relay testing, breaker/fuse coordination. * Auxiliary Feedwater Pumps Discharge Check Valves (2AF-148 and 2AF-107): The inspectors reviewed the IST surveillance results to verify the acceptance criteria were met and to identify any performance degradation. Also, the

inspectors reviewed the pipe stress analysis and pipe support calculations to verify the piping and pipe supports, which support this check valve, meet the design basis requirements. The inspectors reviewed the condition reports and

analyses to ensure the issue was adequately evaluated and corrective actions

were performed or scheduled to address the concern. b. Findings (1) Failure to Monitor Average Outside Temperature Introduction: The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design

Control," involving the licensee's failure to correctly translate design basis assumption

into procedures or instructions. Specifically, the licensee failed to monitor the average

outside air temperature which was one of the design inputs to temperature heat-up calculation associated with rooms that housed vital equipment required during design basis events. Description: Design Basis Calculation 2005-0054, "Control Building GOTHIC Temperature Calculation," evaluated the heat-up rate of various rooms including the

TDAFW pumps room and vital switchgear room. This calculation also determined the

required number of temporary fans needed to maintain the temperature below the maximum allowed. Calculation 2005-0054 used two temperature inputs to the code: (1) maximum outside temperature at a specific time of 95 degrees Fahrenheit (oF); and, (2) maximum outside temperature averaged over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period of 86.6 oF. These temperature inputs were used in the calculation to determine the maximum temperature

in the above mentioned rooms given different accident scenarios including design basis, SBO and Appendix R fire. The maximum outside temperature of 95 oF was used as an input to the calculation in order to bound the most limiting environmental conditions the

station was allowed. The maximum average outside temperature was used as an input

because the calculation was time-dependent and it credited the drop in temperature over

night. Using the average outside temperature allowed the licensee to have a more

accurate calculation in lieu of conservatisms. On August 24, 2011, while reviewing Calculation 2005-0054, the inspectors noticed the licensee was monitoring the maximum outside temperature for 95 oF. The licensee provided instructions to perform a prompt engineering evaluation in the event the

outside temperature exceeded 95 oF to ensure the calculation was still bounded by

12 Enclosure other conservatisms. However, the inspectors noticed the licensee did not monitor the average outside temperature over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period to ensure it did not exceed the

value of 86.6 oF. The inspectors were concerned the failure to monitor the average outside temperature could result in a condition where the temperature in these vital rooms would be outside the design basis calculation. Specifically, the temperature could be below 95 oF, but the average temperature over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period could exceed 86.6 oF. In addition, by the time the maximum temperature of the outside air reaches 95 oF, the average temperature over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period could have already been exceeded. In addition, by not monitoring average outside air temperature over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the licensee would not be able to take adequate compensatory measures to ensure the potential degraded condition does not result in a more significant concern. The licensee acknowledged the inspectors' concerns and initiated corrective action program document AR 01680705 to address the issue. As part of their corrective

actions, the licensee's recommendation included performing an evaluation and additional monitoring once the outside temperature reaches 86.6F. The inspectors reviewed the licensee's action request and had no concerns. In addition, during the licensee apparent cause evaluation (ACE) for this issue, the licensee discovered when the calculation was generated, there was a recommended

action to revise the operator logs, but the action was not implemented. The recommendation was made in an operational decision making (ODM) document. The action was canceled when the ODM document was canceled because licensed

operators incorrectly determined the condition was a functionality, not an operability

issue. Analysis: The inspectors determined the failure to correctly translate the average outside temperature into procedures and instructions were contrary to 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and was a performance deficiency. The performance deficiency was determined to be more than minor because it was

associated with the Mitigating Systems Cornerstone and if left uncorrected, it would have

the potential to lead to a more significant safety concern. Specifically, because the

average outside temperature over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period was not being monitored, the licensee failed to ensure the maximum allowed temperature in the TDAFW pumps room and vital switchgear room would not be exceeded and affect equipment relied upon to perform a safety function during a design basis. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1 -

Initial Screening and Characterization of Findings," Table 4a for the Mitigating System

cornerstone. The finding screened as of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk-significant due to a seismic,

flooding, or severe weather initiating event. Specifically, the licensee provided historical

data showed the average maximum temperature over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period did not exceed 86.6 oF since the calculation was issued. The inspectors determined the finding had a cross-cutting aspect in the area of human performance because the licensee did not ensure adequate training and qualification of

13 Enclosure personnel to ensure nuclear safety. Specifically, the licensee failed to adequately train licensed operators to ensure adequate knowledge with respect to the interface between

functionality of a non-safety system component and the impact of a failure on the

operability of safety-related equipment. H.2(b) Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control" requires, in part, that measures be established to ensure the design basis requirements are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, as of March 24, 2009, the licensee's design control measures failed to verify the design inputs were incorporated into instructions. Specifically, the

licensee failed to monitor average outside air temperature which was an input to a

design basis calculation associated with the TDAFW pumps room and vital switchgear

room temperature heat-up. Because this violation was of very low safety significance and because the issue was entered into the licensee's corrective action program as AR 01680705, this violation is being treated as an NCV, consistent with Section 2.3.2 of

the NRC Enforcement Policy (NCV 05000266/2011009-01; 05000301/2011009-01, Failure to Monitor Outside Air Temperature). .4 Operating Experience a. Inspection Scope The inspectors reviewed 4 operating experience issues to ensure the NRC generic concerns had been adequately evaluated and addressed by the licensee. The operating experience issues listed below were reviewed as part of this inspection: * IN 1987-53, "AFW Pump Trips Resulting from Low Suction Pressure"; * IN 2007-34, "Operating Experience Regarding Electrical Circuit Breakers"; * IN 2006-31, "Inadequate Fault Interrupting Rating of Breakers"; and * GL 89-13, "Service Water System Problems Affecting Safety-Related Systems." b. Findings No findings of significance were identified. .5 Operating Procedure Accident Scenario Reviews a. Inspection Scope The inspectors performed a detailed reviewed of the procedures listed below associated with the Auxiliary Feedwater System. For the procedures listed, the time critical operator

actions were reviewed for reasonableness, in plant actions were walked down with a licensed operator, and any interfaces with other departments were evaluated. The procedures were compared to UFSAR, design assumptions, and training materials to

ensure for constancy. In addition, the inspectors also observed operator actions during

14 Enclosure the performance of four selected scenarios on the station simulator, the station blackout (SBO) event, the anticipated transient without a scram (ATWS) event, the steam

generator tube rupture (SGTR) event, and a faulted steam generator event. The following operating procedures were reviewed in detail: * EOP-0, "Reactor Trip of Safety Injection"; * EOP-0.1, "Reactor Trip Response"; * EOP-1, "Loss of Reactor or Secondary Coolant"; * EOP-1.1, "Safety Injection (SI) Termination"; * EOP-1.2, "Post LOCA Cooldown and Depressurization"; * EOP-2, "Faulted Steam Generator"; * EOP-3, "Steam Generator Tube Rupture"; * EOP-3.1, "Post-SGTR Cooldown using Backfill"; * ECA-0.0, "Loss of All AC Power"; and * CSP-S.1, "Response to Nuclear Power Generation/ATWS." b. Findings (1) Failure to Incorporate Minimum AFW Flow Rate Requirement Into Emergency Procedures Introduction: The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design

Control," involving the licensee's failure to maintain Emergency Procedures consistent

with the Loss of Normal Feedwater (LONF) Accident Analysis. The accident analysis of

record assumed an Auxiliary Feedwater flow rate of 275 gpm, while the inadequate Emergency Procedure allowed the operator to inject AFW flow at a rate greater than 230 gpm, which would allow less than the required amount of 275 gpm of AFW flow. Description: The AFW system was redesigned, in part, to support implementation of the extended power uprate (EPU). The licensee installed one new motor-driven auxiliary

feedwater (MDAFW) pump for each unit in a new location in the auxiliary building. The

pumps, 1P-53 and 2P-53, replaced the safety-related function of the old MDAFW pumps

which had been shared between the two units. The new pumps are unitized, capable of a higher flow capacity, and capable of delivering flow to either or both of the unit's two steam generators (SGs). The new pumps were designed to deliver the minimum flow

requirement of 275 gpm at the lowest SG safety relief valve setpoint. The old AFW

pumps were not removed from the plant, however; they were reclassified as non-safety-

15 Enclosure related pumps and are used during plant start up and shut down. The currently installed safety-related turbine-driven auxiliary feedwater (TDAFW) pumps for each unit meet

EPU design flow requirements, and the new MDAFW pumps will not affect operation of

the TDAFW pumps. In addition, as part of the modification, the licensee installed cavitating venturis in the flow path between the new MDAFW pump to each SG. These venturis were installed as pump runout protection. Specifically, in the event of a failed flow control valve, the venturi would limit the flow from the MDAFW pump to 230 gpm, even while delivering flow to a depressurized SG. The other intact SG would still receive the required flow rate, since the flow rate of 230 gpm would be limited to the faulted SG. The inspectors reviewed the bounding analysis for AFW flow rate requirements; the Loss of Normal Feedwater (LONF)/Loss of AC Power (LOAC) for EPU. This calculation was performed by Westinghouse, as documented in calculation CN-TA-08-79, Revision 1. Here, it was determined the required AFW flow during the LONF event, which bounds

the LOAC event, was 275 gpm, split between the two SGs (137.5 gpm flow split). The calculation concluded the LONF event did not cause any adverse condition in the core,

since it did not result in water relief from neither the pressurizer power operated relief valves, or ASME Code safety valves. The inspectors also reviewed procedure EOP-0.1,"Reactor Trip Response," which would be entered on a LONF event. The procedure was revised as part of EPU, and included a new required AFW flow rate of greater than 230 gpm when the pumps are aligned to

the steam generators. The 230 gpm flow rate was based on the maximum flow rate that

could be delivered to one SG, with only the MDAFW pump available, because of the cavitating venturis installed in the flow path between the new MDAFW pump to each SG. However, in contrast to what was stated in EOP-0.1, the inspectors concluded 275 gpm was required to be delivered to the SGs when both SGs were available during a LONF event. In response to the inspectors' concern, the licensee initiated AR01678638 to revise the EOPs to incorporate the design value for the minimum AFW flow of 275 gpm when supplying both SGs during a LONF event, as specified in the design basis calculations. In addition, the inspectors also reviewed the licensing basis for the new MDAFW pumps discussed in the Safety Evaluation Report (SER) for power uprate. This document stated the new MDAFW pump could deliver 275 gpm to one, or both, SGs (emphasis added) for a steam generator tube rupture event. However, due to the cavitating

venturis installed in the flowpath to each SG, the MDAFW pumps could only deliver a

maximum of 230 gpm to the intact SG, which is in conflict to what was stated in the SER. Upon discussion with NRR technical reviewers, and the licensee, it was determined the SER required a clarification to state the flow to a single SG was limited to 230 gpm when

the MDAFW pump is operating without the TDAFW pump. Additional analysis was

provided to the inspectors which indicated 230 gpm was sufficient flow rate to the intact SG.

16 Enclosure Analysis: The inspectors determined the failure to ensure a minimum AFW flow of 275 gpm as specified in the accident analysis for the Loss of Normal Feedwater event was

contrary to 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and was a performance deficiency. The performance deficiency was associated with the Mitigating System Cornerstone attribute of design control and determined to be more than minor

because if left uncorrected, could become a more significant safety concern.

Specifically, the failure to properly implement the minimum AFW flow rate of 275 gpm

into the Emergency Procedures for the Loss of Normal Feedwater event did not ensure

the pressurizer would not become water solid and cause an over-pressure condition within the Reactor Coolant System during the event. This over-pressure condition may cause liquid water to pass through the Pressurizer Safety Valves which could lead to a more serious Loss of Coolant Accident (LOCA) event. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1 -

Initial Screening and Characterization of Findings," Table 4a for the Mitigating System

cornerstone. The finding screened as of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of safety function, and did not screen as potentially risk-significant due to a seismic, flooding, or

severe weather initiating event. Specifically, although the procedure stated a flow rate

of 230 gpm, the operators could increase flow if needed since the MDAFW pumps were capable of providing greater than 275 gpm to two steam generators if required. The inspectors determined the finding had a cross-cutting aspect in the area of human performance, resources because the licensee failed to ensure the emergency procedures were adequate and included the design basis values. Specifically, the licensee incorporated a non-conservative design value for the minimum AFW flow rate of

230 gpm instead of the design analysis value of 275 gpm specified for LONF event. H.2.c] Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control" requires, in part, that measures shall be established to ensure the applicable regulatory

requirements and the design basis are correctly translated into specifications, drawings, procedures and instructions. A Loss of Normal Feedwater is an analyzed accident in Chapter 14.1.10 of the Point Beach UFSAR. Technical Specification 5.4.1 requires, in

part, that Emergency Procedures will implement the requirements of NUREG-0737.

NUREG-0737 states, in part, that emergency procedures are required to be consistent with the actions necessary to cope with the transients and accidents analyzed. Contrary to the above as of September 2, 2011, the licensee's design control measures failed to correctly incorporate the correct AFW flow rate into the stations emergency operating procedures. Specifically, the accident analysis of record assumes an AFW flow rate of 275 gpm, while the Emergency Procedure allows the operator to inject AFW

flow at a rate "greater than 230 gpm" which would allow less than the required amount

of 275 gpm of AFW flow. Because this violation was of very low safety significance

and because the issue was entered into the licensee's corrective action program as

AR 01678638, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000266/2011009-02; 05000301/2011009-02;

17 Enclosure Failure to Incorporate Minimum AFW Flow Rate Requirement into Emergency Procedures). 4. OTHER ACTIVITIES 4OA2 Identification and Resolution of Problems .1 Review of Items Entered Into the Corrective Action Program a. Inspection Scope The inspectors reviewed a sample of the selected component problems that were identified by the licensee and entered into the corrective action program. The inspectors

reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions related to design issues. In addition, corrective action documents written on issues identified during the inspection were

reviewed to verify adequate problem identification and incorporation of the problem into

the corrective action program. The specific corrective action documents that were

sampled and reviewed by the inspectors are listed in the Attachment to this report. The inspectors also selected 3 issues that were identified during previous CDBIs to verify the concern was adequately evaluated and corrective actions were identified and

implemented to resolve the concern, as necessary. The following issues were reviewed: * NCV 05000266/2008009-01; 05000301/2008009-01, Equalizing Charge Voltage Not Bounded by Battery Room Hydrogen Generation Calculation; * NCV 05000266/2008009-02; 05000301/2008009-02, Non-Conservative Design Basis for Primary Auxiliary Building Heat-up; and * NCV 05000266/2008009-03; 05000301/2008009-03, Ability to Transfer Fuel Oil between EDG Fuel Oil Tanks T-175A/B has not been demonstrated by Testing. b. Findings No findings of significance were identified. 4OA5 Power Uprate (71004) .1 Plant Modifications (2 samples) a. Inspection Scope The inspectors reviewed plant modifications for those implemented for the extended power uprate. This includes seismic qualification of balance of plant piping and pipe

supports for extended power uprate. * Engineering Change EC-12070, "Unit 2 Main Steam and Feedwater pipe support," Revision 0; and

18 Enclosure * EC-11795, "Unit 2 Containment Spray Piping Supports," Revision 0 b. Findings (1) Containment Spray Pipe Support Deficiencies Introduction: The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III,

"Design Control," for failure to meet Seismic Category I requirements for containment

spray piping. Specifically, the licensee failed to provide sufficient justification for the design margin in the Containment Spray Pipe Support 2S-249 and Containment Spray

Pipe Anchor 2A-35 despite the applied bending stress being greater than the allowable

bending stress. Description: The containment spray system per UFSAR Section 6.4.1 has the following safety-related design basis functions: provide sufficient heat removal capability to

maintain the post accident containment pressure below the design pressure, to remove

iodine from the containment atmosphere should it be released in the event of a loss-of-

coolant accident and to provide sufficient sodium hydroxide from spray additive tank to achieve the required sump Ph level in order to prevent chloride induced stress corrosion cracking. The containment spray piping and pipe supports were designed to Seismic Category I requirements as described in UFSAR Section A.5.2. Calculation WE-200074, "Subsystem 6"-SI-301R-1: Containment Spray System from Containment Penetration P-54 to Anchors 2A-34 and 2A-35", Revision 1, evaluated

Containment Spray Pipe Support 2S-249 and Containment Spray Pipe Anchor 2A-35 in

accordance with Seismic Category I requirements for all design basis loading. The pipe

support and pipe anchor support were analyzed to withstand applied stress due to dead loads, live loads, seismic loads, and thermal loads. The inspectors noticed in Calculation WE-200074, Revision 1, Attachment D, the licensee used an allowable

overstress condition, the applied stress was greater than allowable stress, to

demonstrate seismic Category I compliance which was not in accordance with the

design and licensing basis. The Seismic Category I requirements were based on the applied stress less than allowable stress for the evaluation of the Containment Spray Pipe Support 2S-249 and Containment Spray Pipe Anchor 2A-35. The inspectors

determined the use of an allowable overstress condition for Containment Spray Pipe

Support 2S-249 and Containment Spray Pipe Anchor 2A-35 did not meet Seismic Category I requirements. Upon the inspectors' identification of this issue, the license concurred with the inspectors' concern and entered the issue into their corrective action program as

AR01678643, "Overstress of Pipe Supports Analyzed in WE-200074." The licensee performed an additional analysis and determined the pipe support and the pipe anchor were operable but nonconforming. Analysis: The inspectors determined the licensee's failure to meet Seismic Category I requirements for the Containment Spray Pipe Support 2S-249 and Containment Spray

Anchor 2A-35 was contrary to 10 CFR Part 50, Appendix B, Criterion III, "Design

Control," and was a performance deficiency. The performance deficiency was

19 Enclosure determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone

objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, failure to comply with Seismic Category I requirements did not ensure the Containment Spray Pipe Support 2S-249 and

Containment Spray Pipe Anchor 2A-35 would function during a Seismic Category I

design basis event and adversely affect the containment spray piping system and containment barrier. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, "Significance Determination

Process," Attachment 0609.04, "Phase 1-Initial Screening and Characterization of

Findings", Table 4a for Barrier Integrity (Containment Barrier). The finding screened as

of very low safety significance (Green) because the inspectors answered "no" to all four questions in the containment barrier column. Specifically, the licensee was able to show the Containment Spray Pipe Support 2S-249 and Containment Spray Pipe Anchor 2A-35

were operable but nonconforming. The inspectors determined there was no cross-cutting aspect associated with this finding because the deficiency was a legacy design calculational issue and, therefore, was not indicative of licensee's current performance. Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures be established to ensure the applicable regulatory requirements

and the design basis are correctly translated into specifications, drawings, procedures,

and instructions. The design control measures shall provide for verifying or checking the

adequacy of design. Contrary to the above, as of August 17, 2011, the design control measures failed to conform to Seismic Category I requirements and also failed to verify the adequacy of the

design. Specifically, calculation WE-200074 failed to verify the adequacy of the design

for the Containment Spray Pipe Support 2S-249 and Containment Spray Pipe Anchor

2A-35 to ensure it met the Seismic Category I requirements. Because this violation was of very low safety significance (Green) and it was entered into the licensee's corrective action program as AR01678643, this violation is being treated as a Non-Cited Violation,

consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000266/2011009-03; 05000301/2011009-03, Containment Spray Pipe Support Deficiencies). (2) Turbine Building Structural Steel Floor Beams did not meet AISC requirements Introduction: The inspectors identified a finding of very low safety significance (Green) involving the licensee's failure to meet the requirements of American Institute of Steel Construction (AISC) Specifications in the design basis calculation. Specifically, the licensee did not ensure the turbine building structural steel floor beams meet the AISC specifications. No violations of NRC requirements were identified. Description: Design Bases Calculation 12918709-C-0033, "Evaluation of Structural Steel Turbine Building Operating Floor EL. 44 for Change in Pipe Support Reactions,

20 Enclosure Unit 2," Revision 0 evaluated the Turbine Building structural steel floor beams at Elevation 44'-0". The structural steel beams support dead loads, laydown live loads, as

well pipe support loads from the main steam and feedwater piping system which are supported from these beams. The licensee used the American Institute of Steel Construction (AISC) standards to demonstrate structural adequacy of the structural steel

floor beams. Calculation 129187-C-0033 justified, based on engineering judgment, that

a 5 percent overstressed condition of the turbine building structural steel floor beams

was acceptable. Specifically, the licensee stated the maximum interaction ratio (IR)

used for acceptance was less than 1.05. The structure was non-safety-related and the design uses minimum specified yield strength. The actual yield strength of the steel based on mill specification is expected to be higher. The AISC required the allowable stress to be based on the specified minimum yield strength of the material. The licensee used certified material test report strength or

actual material yield strength as a basis for an allowable overstress condition (applied

stress greater than allowable stress) for the evaluation of the turbine building structural

steel floor beams. The use of actual material yield strength as a basis for an allowable overstress condition did not meet the AISC requirements. This issue was entered into the licensee's corrective action program as AR 01682352, "Inadequate Justification for Non-Compliance." Analysis: The inspectors determined the licensee's failure to meet AISC requirements for the turbine building structural steel floor beams was a performance deficiency. The

performance deficiency was determined to be more than minor because the finding was

associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective to limit the likelihood of those events that upset the plant stability and challenge critical safety functions during shutdown, as well

as power operations. Specifically, compliance with AISC requirements for the turbine

building structural steel floor beams ensures the main steam and feedwater piping

system would not be affected during a design basis event. The failure to comply could

impact the piping systems and potentially result in a turbine trip/reactor trip. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase I-Initial Screening and Characterization of

Findings," Table 4a for Initiating Events. The finding screened as of very low safety

significance (Green) because the transient initiator would not contribute to both the

likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The inspectors determined this finding had a cross-cutting aspect in the area of human performance, work practices because the licensee did not ensure effective supervisory and management oversight of work activities, including contractors, such that nuclear

safety was supported. Specifically, the licensee failed to have adequate oversight of

design calculation and documentation for establishing structural adequacy of the turbine building structural steel beams at EL. 44'-0". H.4(c) Enforcement: Since the equipment involved with the performance deficiency were not safety-related, there were no violations of NRC regulations associated with this finding

21 Enclosure (FIN) and as such, no enforcement. (FIN 05000266/2011009-04; 05000301/2011009-04, Turbine Building Structural Steel Floor Beams did not meet AISC requirements) 4OA6 Meeting(s) .1 Exit Meeting Summary On September 2, 2011, the inspectors presented the inspection results to Mr. T. Vehec, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors asked the licensee whether any materials examined during

the inspection should be considered proprietary. Several documents reviewed by the

inspectors were considered proprietary information and were either returned to the licensee or handled in accordance with NRC policy on proprietary information. 4OA7 Licensee-Identified Violations The following violation of very low safety significance (Green) was identified by the licensee and was a violation of NRC requirements, which meets the criteria of

Section VI.A.1 of the NRC Enforcement Policy for being dispositioned as an NCV. * A finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," was identified by the licensee for the failure to ensure adequate instructions were adequately prescribed in procedures. Specifically, the licensee failed to ensure the

receptacle 2PR-49 listed in Procedure AOP-30, "Temporary Ventilation for Vital

Areas," as one of the three potential power sources for transformer X-71 adequate

for the transformer plug, was acceptable, in that the receptacle and transformer had difference phase connections. This transformer would be used to power temporary fans relied upon for design basis accident and the loss of the normal/fixed

ventilations in the AFW and switchgear rooms. The performance deficiency was

determined to be more than minor because it was associated with the Mitigating

Systems Cornerstone attribute of Equipment Performance, and affected the

cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The SDP Phase I evaluation concluded the finding screened as of very low safety significance. This issue was entered into the licensee's corrective action as AR01652555, as a

corrective action, the licensee prepared an EC 271778 to modify the receptacle

during the next Unit Refueling Outage. The inspectors also noticed procedure AOP-30 still showed 2PR-49 as one of the potential power sources. The inspectors were concerned there were no compensatory measures in place identifying that this power

source could not be used and also identifying other receptacles in the area that could

be utilized as an interim measure. The licensee entered the inspectors concern into

their corrective action program as AR01682644. ATTACHMENT: SUPPLEMENTAL INFORMATION

1 Attachment SUPPLEMENTAL INFORMATION KEY POINTS OF CONTACT Licensee T. Vehec, Plant General Manager J. Atkins, Operational Assistant Manager

S. Brown, Program Engineering Manager

L. Bruster, Engineering

D. Craine, Radiation Protection Manager

F. Flentje, Licensing Supervisor V. Kanal, Engineering Supervisor T. Kendall, Engineering

J. Kenney, Mechanical Department

J. Lewandowski, Quality Assurance Supervisor

T. Lensmire, Electrical Design Engineering A. Mitchell, Performance Improvement Manager M. Moran, EPU Engineering manager

L. Nicholson, Licensing Director

J. Pierce, Training Assistant Manager

B. Scherwinski, Licensing P. Wild, Design Engineering Manager B. Woyak, Engineering Supervisor

Nuclear Regulatory Commission S. Burton, Senior Resident Inspector M. Thorpe-Kavanaugh, Resident Inspector

Attachment 2LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened and Closed 05000266/2011009-01; 05000301/2011009-01 NCV Failure to Monitor outside Air Temperature (Section 1R21.3.b (1))05000266/2011009-02; 05000301/2011009-02 NCV Failure to Incorporate Minimum AFW Flow Requirement into Emergency Procedures (Section 1R21.6.b (1))05000266/2011009-03; 05000301/2011009-03 NCV Containment Spray Pipe Support Deficiencies (Section 4OA5.1.b (1))05000266/2011009-04; 05000301/2011009-04 FIN Turbine Building Structural Steel Floor Beams Did Not Meet AISC Requirements (Section 4OA5.1.b (2))

Attachment 3LIST OF DOCUMENTS REVIEWED The following is a list of documents reviewed during the inspection. Inclusion on this list does not imply the NRC inspectors reviewed the documents in their entirety, but rather, that selected sections of portions of the documents were evaluated as part of the overall inspection effort. Inclusion of a document on this list does not imply NRC acceptance of the document or any part of it, unless this is stated in the body of the inspection report. CALCULATIONS Number Description or Title RevisionN-93-057 Battery D-06 DC System Sizing, Voltage Drop, and Short Circuit Calculations 6 N-93-041 Hydrogen buildup in the Battery Rooms 3 2003-046 Battery Chargers Sizing and Current Limit Set Point 4 P-94-004 MOV Overload Heater Evaluation 13 P-94-004 MOV Overload Heater Evaluation 13C P-89-031 Voltage Drop Across MOV Power Lines 12 N-98-095 Minimum DC Control Voltage Available at CC and TC of Circuit Breakers at 4160 Safety Switchgears and 480 Safety

Load Centers 3 2009-0027 Cable Ampacity and Voltage Drop for DC Power Cables 0 N-92-005 125 VDC Coordination Analysis 2A P-90-017 Motor Operated Valve Undervoltage Stem Thrust and Torque 22 97-0231 Auxiliary Feedwater Pump Low Suction Pressure SW Switchover and Pump Trip Instrument Loop

Uncertainty/Setpoint Calculation 2 97-0231 Auxiliary Feedwater Pump Low Suction Pressure SW Switchover and Pump Trip Instrument Loop

Uncertainty/Setpoint Calculation 002-B PBNP-IC-42 Condensate Storage Tank Water Level Instrument Scaling and Loop Uncertainty/Setpoint Calculation Rev 002-A 2008-0024 AFWP Room Flood Basis Calculation Rev 0 2010-0022 Flow Parameter EOP Setpoints Calculation Rev 0 2005-0008 Minimum Voltage Requirements for SR MCC Control Circuits 0 P-94-004 MOV Overload Heater Evaluation 13 & 13C2004-0009 13.8KV and 4.16KV Protection and Coordination 2-N P-90-017 MOV UV Stem Thrust and Torque Calculation 22 P-89-031 Voltage Drop Across MOV Power Lines 12 2001-0033 Electrical Input Calc, 345kV - 480V SWGR Circuits 9 2001-0049 480V Switchgear Coordination and Protection 2 2004-0001 AC Electrical System Analysis - Model Inputs 9 2004-0002 AC Electrical System Analysis 4 2008-0014 Determination of Power Cable Ampacities and Verification of Overload Protection 0 2005-0007 Electrical System Transient Analysis 3

Attachment 4CALCULATIONS Number Description or Title RevisionN-94-007 MOV Motor Brake Voltage Evaluation 0 2008-0005 4160/480V Loss of Voltage and Under-Frequency Relay Settings 2 2003-0014 MOV Operating Parameters 6 2005-0053 Primary Aux Building GOTHIC Temperature Calculation 0 2009-06020 Maximum Allowable Working Pressure and Evaluation of Valves and Components of the AFW System 1 2009-08450 AFW Air Operated Valves Component Level Calculation 0 2009-06929 AFW Air Operated Valves Functional and MEDP Calculation 0 2009-06932 Nitrogen or Compressed Air Backup System for MDAFP (1,2-P53) Discharge Valves and Flow Recirc. Valves 1 P-94-005 MOV Stem Thrust Calculation 11 97-0231 AFW Pump Low Suction Pressure SW Switchover and Pump Trip Inst. Loop Uncertainty/Setpoint Calc 2 2010-0010 AFW Low-Low-Low SW Switchover Instrument Loop Unc/Setpoint Calc., 0 WEP-SPT-33 AFW Flow Indication Uncertainty 4 CN-CPS-07-6 Point Beach S/G Narrow Range Level Instr. Uncertainty and Setpoint Calc. as Modified to Reflect Operations at Pre EPU and Post EPU Conditions (IC-25) 3 CN-TA-08-79 Point Beach Units 1 and 2 Loss of Normal Feedwater/Loss of AC Power (LONF/LOAC) Analysis for the EPU Program 1 CN-CRA-08-40 SGTR Thermal Hydraulic Input to Dose Analysis for Point Beach Units 1 and 2 to Support EPU 0 CN-CRA-08-10 Point Beach EPU Steam Line Break Inside Containment Mass/Energy Release 1 2003-0062 AFW Pump NPSH Calculation and CST Volume Required to Prevent Vortexing 2-B 2009-06582 Available Water in Volume of Piping in Protected Portion of MDAFW Pump Suction 0 S-11165-116-05 AFW Pump Anchorage Design and Foundation Analysis 1 96-0244 Minimum Allowable IST Acceptance Criteria for TDAFW and MDAFW Pump Performance 3 N-94-019 Determination of Conditions for MOV Pressure Locking and Thermal Binding 000-B 2005-0054 Control Building GOTHIC Temperature Calculation 1 WE-300089 MDAFW Pump Suction Piping from CSTs T-24A and T-24B to Anchor 0 WE-300090 MDAFW Common Recirculation Piping from CST to Anchor HD-8-026-3A 00-A WE-300089 MDAFW Common Suction Piping from CST's to Anchor HD-8-049-3A 00-A

Attachment 5CALCULATIONS Number Description or Title RevisionWE-200052 Auxiliary Feedwater System from Structural Anchors DB3-2H7 and DB3-2H4 to Containment Penetration P5 (EB10-A13) 00-B/C/D WE-200051S Emergency FW from Penet. P-5 & 6 to Anchors H-11, 2H2, 2H4 & 2H7 00-C S-11165-116-07 Pipe Support Qualification for AFW Margin Improvements 1 129187-P-0011 Unit 2, Main Steam outside Containment - Piping Qualification for Extended Power Uprate Conditions 6 129187-P-0018 Unit 2, Fedwater outside Containment - Piping Qualification for Extended Power Uprate Conditions 6 PBNP-994-21-06 HELB Reconstitution Program - Task 6 Break and Crack Size/Location Selection 2 129187-C-0055 Evaluation of Main Steam Pipe Supporting Structure of Unit #2 Façade and Turbine Buildings for Changes in Pipe Support Reactions Associated with Uprate Conditions (EC-12070) 0 129187-C-0054 Evaluations of Pipe Supporting Structures of Unit #2 Auxiliary Building for Changes in Pipe Support Reactions Associated

with Uprate Conditions 0 12918709-C-0052 Evaluation of Main Steam and Feedwater Pipe Supporting Structures of Unit 2 Containment Building for Changes in

Pipe Support Reactions 0 12918709-C-0033 Evaluation of Structural Steel Turbine Building Operating Floor EL. 44 for Change in Pipe Support Reactions, Unit 2 0 129187-C-0080 Corrective Action Report of Structural Steel Turbine Building Operating Floor EL. 44 for Legacy Issue, Unit 2 0 WE-200074 Subsystem 6"-SI-301R-1: Containment Spray System from Containment Penetration P-54 to Anchors 2A-34 and 2A-35 1 WE-300048 Subsystem AC-601R/SI-151R: Suction Piping from RWST to SI, CS and RHR 0-H WE-200040 Containment Spray Pump 2-P14A Discharge to P-54 0-A WE-200074 Subsystem 6"-SI-301R-1: Containment Spray System from Containment Penetration P-54 to Anchors 2A-34 and 2A-35 1-C WE-200104 Subsystem AC-601R/SI-151R: Suction Piping from RWST to Safety Injection, Containment Spray and RHR Pumps 0-F WE-200073 Subsystem 6"-SI-301R-1: Containment Spray System from Containment Penetration P-55 to Anchors 2A-36 and 2A-37 1-C WE-100092 Containment Spray System Line 3"-SI-301R-1 between Anchors 1A-34 and 1A-35 0-A WE-100093 Subsystem 6"-SI-301R-1-9: Containment Spray System from Containment Penetration P-55 to Anchors 1A-34 and 1A-35 0-D

Attachment 6CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION Number Description or Title Date AR01674251 Anti-Sweat Insulation Found Removed 8/02/11 AR01674327 Fire Hose Staged Between CSTs for Unknown Activity 8/02/11 AR01674473 OM 3.27 to NP 1.9.6 Process to Process GAP 8/03/11 AR01674481 No Temporary Information Tag on Cubical 2B2-427M AR01674616 Miscellaneous Parts Attached to Body of 2AF-4073 8/03/11 AR01674696 Error Identified in Calculation N-93-057 8/03/11 AR01674699 Damaged Wiring in Plant for Excessively Long Time 8/03/11 AR01674726 NRC Comments on AR Operability Screening 8/03/11 AR01674739 PBNP Response to Prairie Island OE32688 8/03/11 AR01674806 TSB 3.7.5 Potential Changes During FSAR Revisions 8/04/11 AR01675019 Temporary Storage Tag Missing 8/04/11 AR01675023 During a Wlakdown with CDBI NRC Inspectors, Noted two Instances That are in Question AR01675066 RMP 9353 Question by NRC 8/04/11 AR01675074 Emergency Lighting 8/04/11 AR01675094 D-105 Intertier Connection Cable Bend Radios 8/04/11 AR01675253 CL-13E Part 2 Inconsistencies 8/05/11 AR01675812 CL 13E Part2 AFW Valve Lineup Motor Drive 8/08/11 AR01676059 125 Vdc Fuse Issue 8/08/11 AR01677153 Calculation for Vital 120 Vac System 8/11/11 AR01677805 Error in Control Circuit Voltage Drop 8/15/11 AR01677914 Inadequate Documentation of Containment Dome Truss 8/15/11 AR01678123 Lack of Basis Documented in Calculation 2004-0002 8/16/11 AR01678283 2SAF-4000 Thermal Overload Testing 8/16/11 AR01678285 Preventive Maintenance for 2SAF-4000 8/16/11 AR01678535 Discrepancy in 125 Vdc Drawing 8/17/11 AR01678638 Evaluate ERG Setpoint Deviation for AFW Flow Setpoint in EOP 8/17/11 AR01678643 Overstress of Pipe Support Analyzed in WE-200074 8/17/11 AR01679081 New EOP Setpoint for AFW Flow During LONF/LOCA Events 8/18/11 AR01679387 IT 08A and IT 09A Note Require Update 8/19/11 AR01679408 CR for Tracking Priority 1 PCR 01678831 Unit 2 8/19/11 AR01679412 CR for Tracking Priority 1 PCR 01678829 Unit 1 8/19/11 AR01679758 Issue Identified in Calculation P-94-004 8/22/11 AR01679907 ARB C01 A-2-9 T-24A/B Condensate Storage Tanks Level 8/22/11 AR01680185 TLB 34 Condensate Storage Tank T-24A/B 8/23/11 AR01680201 ICP 13.009-2 Condensate Storage Tank Loop Instrument 18 Months 8/23/11 AR01680705 Need to Add Operator Action to Logs 8/24/11 AR01680951 Possible Error Trap in Calculations 8/25/11 AR01681176 CST Low Level Alarm Setpoint have Procedure Issues 8/25/11 AR01681178 Incorrect Snubber Capacity used in EPU Calculation 8/25/11

Attachment 7CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION Number Description or Title Date AR01682352 Inadequate Justification for non-compliance 8/30/11 AR01682644 Issues Identified with AOP-30 8/31/11 AR01682729 Process Issues with Procedure Changes for CST Level Setpoint 8/31/11 CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION Number Description or Title Date AR 01232138 Comments on 125VDC Vendor Calc.'s After Owners Review 08/12/03 AR 01311121 Equipment Outside Short Circuit Rating 01/19/07 AR 01394317 2010 NRC URI-Inverter Transfers to Alt Power During Test 08/07/10 AR01612401 480V SWGR Coordination Recommended Settings not implemented AR01334024 IN 2007-34 Review for applicability 12/17/07 AR01315278 IN 2006-31 Review for applicability 04/04/07 AR01347091 LOV relays may trip during grid faults AR01657810 2B-04 Was De-energized on overcurrent AR01281343 Calculated SC Exceed Equipment Ratings and Capabilities AR01281432 Potential Protective Device Tripping for LOCA with degraded voltage AR01047353 2006 CDBI Violation - OPR153 did not address Seismic event for identified condition AR01303493 2006 CDBI Violation - Calculated SC exceeds equipment ratings 09/21/06 AR01302261 2006 CDBI Violation - Calculated SC exceeds equipment ratings 08/30/06 AR01226467 Cable Overload Protection for existing design not documented AR01331133 Cable Overload Commitments AR01366948 1P-29 TDAFP Outboard Bearing Reached Alert Alarm 06/15/09 AR01371971 1P-29 Turbine Outboard Bearing Temp High 09/15/09 AR01379586 1P-29 TDAFW Pump Outboard Turbine bearing Temp High 01/04/10 AR01392619 1P-29 Turbine Outboard Bearing High Temp Alarm 07/12/10 AR01397577 Engineering Evaluation for 1P-29 Temperature Alert 10/04/10 AR01607140 1TR-2000B PT 19 1P-29 Temperature High Alarm 01/10/11 AR01652555 Test Cables in CSR and 2PR-49 Usability Issue 05/17/11 AR01661563 Pump Secured Due to Outbrd Turb Bearing Temp > 250 Degrees F 06/16/11 AR01669101 Potential Overstresses Beams at EL. 26' of U2 Turbine Building 7/13/11 AR01402167 Calculation 12918709-C-0033 Rev. 1 Existing Conditions 12/21/10

Attachment 8DRAWINGS Number Description or Title Revision 6118 E-6, Sheet 1 125V DC Dist. System 55 6118 E-6, Sheet 2 125 V DC System 19 499B4676, Sheet 840 Elementary Wiring Diagram Turbine Driven Auxiliary Feedwater Pump Discharge Valve 2AF-4001 01 499B466, Sheet 863 Elementary Wiring Diagram 2P-29 Auxiliary Feed Pump Suction from Service Water Supply 14 499B466, Sheet 867 Elementary Wiring Diagram Turbine Driven Auxiliary Feedwater Pump Discharge Valve 2AF-4000 15 499B466, Sheet 1803 Elementary Wiring Diagram Condensate Storage Tank AFW Suction Valve Control 00 499B466, Sheet 899 Elementary Wiring Diagram 2P-053 AFW Pump Service Water Suction Valve 2AF-4067 00 499B466, Sheet 744 Elementary Wiring Diagram Turbine Driven Auxiliary Feedwater Trip/Throttle Valve 2Ms-02082 06 62550 CD2-15-1 Connection Diagram Rack 2C173B-F/2C-197 02 6118 M-2217 P&ID Auxiliary Feedwater System 02 6118 M-217, Sh 1 P&ID Auxiliary Feedwater System 94 6118 M-217, Sh 2 P&ID Auxiliary Feedwater System 25 E-98, Sheet 50D Panel Schedule 125V DC Panel D-28 (D-40) 12 6704-D-323115 Schematic Diagram 4160V Swgr Bus 1-A06 (2-A06) Output Breaker 1A52-86 (2A52-87) from Diesel

Generator G-04 (G-03) 13 6704-D-323101 Schematic Diagram 4160V Swgr Bus 1-A06 (2-A06) Output Breaker 1A52-80 (2A52-93) from Diesel

Generator G-03 (G-04) 15 EPB02EAPW12800209 Three Line Diagram - 2A06 and EDG G-04 9 EPB02EAPK00000130 480V One Line Diagram, 2B03/2B04 30 EPB01EAPS24000108 Schematic 4160V 1A05 8 EPB02EAPK24000112 Schematic 4160V 2A05 12 EPB02EAPK16600215 One Line Diagram MCC 2B42 11 PB07322 Simplified Electrical Power Distribution Single Line 1 PB07322 Simplified Electrical Power Distribution 1 018995 P&ID Service Water 77 019016 P&ID Auxiliary Feedwater System 94 275460 P&ID Auxiliary Feedwater System 20

Attachment 9MISCELLANEOUS Number Description or Title Date or Revision WO 00370104 DC Starter Verification & TOL Test for 2SMS-2019, 2SAF-4001 and 2SAF-4006 04/10/2011 WO 40061953-01 ICP 6.6 Service Water Instrumentation - Controlled WO 40061953-02 ICP 6.6 Service Water Instrumentation - Clean Side 345KV System Health Report 06/30/11 U1/2 4160V System Health Report 06/30/11 U1/2 480V System Health Report 06/30/11 OPR00153 Calculated SC currents exceed equipment ratings 1 DBD-22 Design Basis Document - 4160VAC System 5 DBD-21 Design Basis Document - 480VAC System 5 SE 2008-021 Creation of Procedures for Supplemental Ventilation 04/03/09 Spec No. 6118-M-37 Turbine Building Feed Water Pump Room Ventilation Unit (Stand By) W-46 1 MODIFICATIONS Number Description or Title Date or Revision EC 16640 MOV Capacity during LOOP/LOCA 0 MR 02-039* A/B Aux Feed Water Pump 2-29 Recirculation Line Orifice 03/08/03 EC 12070 Unit 2 Main Steam and Feedwater Pipe Supports 0 EC 11795 Unit 2 Containment Spray Piping Supports 0

Attachment 10PROCEDURES Number Description or Title Revision RMP 9046-2 Station Battery Individual Cell Charging 13 NP 8.4.13 Fuse Replacement 8 2ICP 04.003-5 Auxiliary Feedwater Flow and Pressure Instruments Outage Calibration 16 2ICP 02.031 2P-53 Motor Driven Auxiliary Feedwater Suction Header Pressure Trip Channel Operability Test 0 AOP-13C Severe Weather Conditions Rev 22 ICP06.006 Service Water System Non-Outage Instruments Calibrations Rev 11 NP 5.2.6 FSAR Maintenance Rev 14 NP 5.2.15 Technical Specification Bases Control Rev 11 FP-E-MOD-03 Temporary Modifications Rev 9 BG-ECA-2.1 Uncontrolled Depressuratization of Both Steam Generators Rev 33 2ICP 02.031 2P-53 Motor Driven Auxiliary Feedwater Suction Header Pressure Trip Channel Operability Test Rev 0 TLB 34 Tank Level Book - Condensate Storage Tank T-24 Rev 9 2RMP 9133 Motor Driven and Turbine Drive Auxiliary Feedwater Pump Start on Bus A-01 and A-02 Undervoltage Refuel Calibration Rev 15 STPT 25.1 Emergency Operating Procedure (EOP) Setpoints Rev 4 NP 1.9.6 Plant Cleanliness and Storage Rev 36 ORT 3C Auxiliary Feedwater System and AMSAC Actuation Unit 2 Rev 16 TS 87 Primary Auxiliary Building Ventilation System Monthly Checks Rev 2 STPT 14.11 Auxiliary Feedwater Setpoint Document Rev 23 EOP-0 Reactor Trip of Safety Injection EOP-0.1 Reactor Trip Response Rev 38 EOP-1 Loss of Reactor or Secondary Coolant EOP-1.1 SI Termination EOP-1.2 Post LOCA Cooldown and Depressurization EOP-2 Faulted Steam Generator EOP-3 Steam Generator Tube Rupture EOP-3.1 Post-SGTR Cooldown using Backfill ECA-0.0 Loss of All AC Power Rev 56 ECA-1.1 Loss of Emergency Coolant Recirculation ECA-1.2 LOCA Outside Containment ECA-1.3 Containment Sump Blockage CSP-S.1 Response to Nuclear Power Generation/ATWS AOP-10A Safe Shutdown - Local Control RMP 9366 50VCP-WR350 4.16KV Vacuum Breaker Routine Maintenance 18

Attachment 11PROCEDURES Number Description or Title Revision RMP 9353 ABB 5-HK-350 4.16KV Breaker Routine Maintenance 13 RMP 9374-5 Molded Case Circuit Breaker Testing 5 RMP 9369-1 Westector/Amptector Overload Setpoint Check LV Breakers 21 RMP 9303 Westinghouse DB-50 Breaker Routine Maintenance 23 RMP 9305 Westinghouse DB-75 Breaker Routine Maintenance 20 2ICP 02.032 2P-29 Auxiliary Feedwater Suction Header Pressure Trip Channel Operability Test 0 AOP-10 Control Room Inaccessibility 6 AOP-30 Temporary Ventilation for Vital Areas 7 ARP 2C04 2C 4-4 2TR-2000A or B Temperature Monitor Unit 2 7 STPT 14.11 Setpoint Document Auxiliary Feed Water General Instrumentation Channels 23 SURVEILLANCES (COMPLETED) Number Description or Title Date WO 00370423 Loop 2PT-4069 Functional Check 04/20/2011 RMP 9200-2 Station Battery D-06 Discharge Tests, Recovery and Equalizing Charge 03/24/2009 WO 40066812 125V Station Tech Spec Batteries Weekly Inspection 07/12/2011 WO 40066815 125V Station Tech Spec Batteries Weekly Inspection 08/12/2011 WO 40066814 125V Station Tech Spec Batteries Weekly Inspection 07/26/2011 WO 00390946 D-06, Quarterly Station Battery Inspection 01/10/2011 WO 00384768 D-06, Quarterly Station Battery Inspection 04/12/2011 WO 00395882 D-06, Quarterly Station Battery Inspection per RMP 9046-1 06/21/2011 WO 00368194 D-06, Annual Station Battery Inspection per RMP 9046-1 05/17/2010 WO 00358159 D-06, Annual Station Battery Inspection per RMP 9046-1 05/04/2009 WO 00395879 D-06, Annual Station Battery Inspection per RMP 9046-1 06/21/2011 RMP 9359-5B D-06 Station Battery, D-08 Battery Charger Maintenance and Surveillances 05/04/2009 RMP 9359-5B 125V Station Tech Spec Batteries Weekly Inspection 07/30/2010 WO 0366265 D-06 Modified Performance Test 05/04/2009 WO 00384765 D-06, Station Battery Service Test 01/06/2010 2ICP 02.031 2P-53 Motor Driven Auxiliary Feedwater Suction Header pressure Trip Channel Operability Test 08/16/110 IT 09A Cold Start of Turbine-Driven Auxiliary Feed Pump and Valve Test (Quarterly) Unit 2 02/15/11 IT 09A Cold Start of Turbine-Driven Auxiliary Feed Pump and Valve Test (Quarterly) Unit 2 06/16/11 PC 75 Part 8 AOP Fan and Air Compressor Surveillance Test 05/14/10

Attachment 12SURVEILLANCES (COMPLETED) Number Description or Title Date ORT 59 Operations Refueling Test for Unit 1 and 2 Train A Spray System CIV Leakage Test ORT 60 Operations Refueling Test for Unit 1 and 2 Train B Spray System CIV Leakage Test IT 05 Inservice Test for Unit 1 Train A and B Containment Spray Pump and Valves IT 06 Inservice Test for Unit 2 Train A and B Containment Spray Pump and Valves WORK DOCUMENTS Number Description or Title Date 380449 01 2X-14 Obtain Oil Sample for Dissolved Gas 03/24/11 380477 01 2B-42 MCCB Primary Current Injection Testing 03/21/11 333020 01 A52-HK-1200-08 Breaker Maintenance Per RMP 9353 02/18/08 378410 01 B52-DB50-006 Breaker Maintenance Per RMP 9303 (Feeder Bkr) 11/09/10 359726 01 B52-DB75-004 Breaker Maintenance Per RMP 9305 (Supply Bkr) 06/07/11 382090 01 4160V A-05 SWGR Infrared Survey 02/15/11 392343 01 4160V A-06 SWGR Infrared Survey 02/09/11

Attachment 13LIST OF ACRONYMS USED AC Alternating Current ACE Apparent Cause Evaluation ADAMS Agencywide Document Access Management System AFW Auxiliary Feedwater

AOP Abnormal Operating Procedure

AR Action Request

AISC American Institute of Steal Construction

ASME American Society of Mechanical Engineers CDBI Component Design Bases Inspection CFR Code of Federal Regulations

CST Condensate Storage Tank

DRS Division of Reactor Safety

EOP Emergency Operating Procedure EPU Extended Power Uprate °F Fahrenheit Degrees

FIN Finding

GL Generic Letter

IMC Inspection Manual Chapter IN Information Notice IR Inspection Report

IST Inservice Testing

kV Kilovolt

LOCA Loss of Coolant Accident

LONF Loss of Normal Feedwater LOOP Loss of Off-site Power MDAFW Motor Driven Auxiliary Feedwater

MOV Motor-Operated Valve

NCV Non-Cited Violation

NPSH Net Positive Suction Head NRC U.S. Nuclear Regulatory Commission ODM Operational Decision Making

OM Operation and Maintenance

PARS Publicly Available Records System

psig Pressure Per Square Inch Gage

RIS Regulatory Issue Summary SBO Station Blackout SDP Significance Determination Process

TDAFW Turbine Driven Auxiliary Feedwater

TS Technical Specification

UFSAR Updated Final Safety Analysis Report VAC Volts Alternating Current VDC Volts Direct Current

L. Meyer -2- 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 System (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). Sincerely, Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-266; 50-301 License No. DPR-24; DPR-27 Enclosure: Inspection Report 05000266/2011009; 05000301/2011009 w/Attachment: Supplemental Information cc w/encl: Distribution via ListServ DISTRIBUTION: Daniel Merzke

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Cynthia Pederson Steven Orth Jared Heck

Allan Barker Carole Ariano Linda Linn

DRPIII DRSIII Patricia Buckley Tammy Tomczak ROPreports Resource DOCUMENT NAME: G:\DRSIII\DRS\Work in Progress\-PTBCH 2011 009 CDBI AKD.docx Publicly Available Non-Publicly Available Sensitive Non-Sensitive To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy OFFICE RIII RIII NAME ADahbur:ls AMStone DATE 10/17/11 10/17/11 OFFICIAL RECORD COPY