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:	Ann Marie Stone 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 Commissions rules and regulations and with the conditions of your license.

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

personnel.

Based on the results of this inspection, 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 NRCs

program for overseeing the safe operation of commercial nuclear power reactors is described in

NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

A.

NRC-Identified and Self-Revealed Findings

Cornerstone: Initiating Events

Green. The inspectors identified a finding of very low safety significance involving the

licensees failure to meet the requirements of the American Institute of Steel

Construction (AISC) Specification. Specifically, the licensees design basis calculation

failed to ensure the turbine building structural steel floor beams met the AISC

specification. This finding was entered into the licensees 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 plants 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 licensees 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 licensees 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 licensees 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 licensees 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 licensees 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 licensees 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 licensees PRA, the Point Beachs 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.

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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 switchgears 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 stations 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 licensees 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 licensees 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

switchgears 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 licensees 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

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Enclosure

MCCs 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 licensees 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

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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 licensees 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

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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/licensees 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

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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 stations 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

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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 licensees 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 licensees recommendation included performing an evaluation and

additional monitoring once the outside temperature reaches 86.6F. The inspectors

reviewed the licensees 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 licensees 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 licensees 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 licensees 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 units 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 licensees 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 licensees 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 licensees 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 licensees 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 licensees 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 licensees 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 licensees corrective action program as AR 01682352, Inadequate Justification for

Non-Compliance.

Analysis: The inspectors determined the licensees 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 licensees 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

2

LIST 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

3

LIST 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

Revision

N-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 & 13C

2004-0009

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

4

CALCULATIONS

Number

Description or Title

Revision

N-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

5

CALCULATIONS

Number

Description or Title

Revision

WE-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

1. 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

6

CORRECTIVE 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

7

CORRECTIVE 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

8

DRAWINGS

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

EPB02EAPW128002

09

Three Line Diagram - 2A06 and EDG G-04

9

EPB02EAPK0000013

0

480V One Line Diagram, 2B03/2B04

30

EPB01EAPS2400010

8

Schematic 4160V 1A05

8

EPB02EAPK2400011

2

Schematic 4160V 2A05

12

EPB02EAPK1660021

5

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

9

MISCELLANEOUS

Number

Description or Title

Date or

Revision

WO 00370104

DC Starter Verification & TOL Test for 2SMS-2019,

2SAF-4001 and 2SAF-4006

04/10/20

11

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

10

PROCEDURES

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

11

PROCEDURES

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

12

SURVEILLANCES (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

13

LIST 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

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DATE

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