ML11291A094
| 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.
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:
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.
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
Failure to Monitor outside Air Temperature (Section
1R21.3.b (1))05000266/2011009-02; 05000301/2011009-02
Failure to Incorporate Minimum AFW Flow Requirement
into Emergency Procedures (Section 1R21.6.b (1))05000266/2011009-03; 05000301/2011009-03
Containment Spray Pipe Support Deficiencies (Section
4OA5.1.b (1))05000266/2011009-04; 05000301/2011009-04
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
- 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
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
0-A
WE-100093
Subsystem 6-SI-301R-1-9: Containment Spray System
from Containment Penetration P-55 to Anchors 1A-34 and
0-D
Attachment
6
CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION
Number
Description or Title
Date
Anti-Sweat Insulation Found Removed
8/02/11
Fire Hose Staged Between CSTs for Unknown Activity
8/02/11
OM 3.27 to NP 1.9.6 Process to Process GAP
8/03/11
No Temporary Information Tag on Cubical 2B2-427M
Miscellaneous Parts Attached to Body of 2AF-4073
8/03/11
Error Identified in Calculation N-93-057
8/03/11
Damaged Wiring in Plant for Excessively Long Time
8/03/11
NRC Comments on AR Operability Screening
8/03/11
PBNP Response to Prairie Island OE32688
8/03/11
TSB 3.7.5 Potential Changes During FSAR Revisions
8/04/11
Temporary Storage Tag Missing
8/04/11
During a Wlakdown with CDBI NRC Inspectors, Noted two
Instances That are in Question
RMP 9353 Question by NRC
8/04/11
8/04/11
D-105 Intertier Connection Cable Bend Radios
8/04/11
CL-13E Part 2 Inconsistencies
8/05/11
CL 13E Part2 AFW Valve Lineup Motor Drive
8/08/11
125 Vdc Fuse Issue
8/08/11
Calculation for Vital 120 Vac System
8/11/11
Error in Control Circuit Voltage Drop
8/15/11
Inadequate Documentation of Containment Dome Truss
8/15/11
Lack of Basis Documented in Calculation 2004-0002
8/16/11
2SAF-4000 Thermal Overload Testing
8/16/11
Preventive Maintenance for 2SAF-4000
8/16/11
Discrepancy in 125 Vdc Drawing
8/17/11
Evaluate ERG Setpoint Deviation for AFW Flow Setpoint in
8/17/11
Overstress of Pipe Support Analyzed in WE-200074
8/17/11
New EOP Setpoint for AFW Flow During LONF/LOCA Events
8/18/11
IT 08A and IT 09A Note Require Update
8/19/11
CR for Tracking Priority 1 PCR 01678831 Unit 2
8/19/11
CR for Tracking Priority 1 PCR 01678829 Unit 1
8/19/11
Issue Identified in Calculation P-94-004
8/22/11
ARB C01 A-2-9 T-24A/B Condensate Storage Tanks Level
8/22/11
TLB 34 Condensate Storage Tank T-24A/B
8/23/11
ICP 13.009-2 Condensate Storage Tank Loop Instrument 18
Months
8/23/11
Need to Add Operator Action to Logs
8/24/11
Possible Error Trap in Calculations
8/25/11
CST Low Level Alarm Setpoint have Procedure Issues
8/25/11
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
Inadequate Justification for non-compliance
8/30/11
Issues Identified with AOP-30
8/31/11
Process Issues with Procedure Changes for CST Level
Setpoint
8/31/11
CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION
Number
Description or Title
Date
Comments on 125VDC Vendor Calc.s After Owners Review
08/12/03
Equipment Outside Short Circuit Rating
01/19/07
2010 NRC URI-Inverter Transfers to Alt Power During Test
08/07/10
480V SWGR Coordination Recommended Settings
not implemented
IN 2007-34 Review for applicability
12/17/07
IN 2006-31 Review for applicability
04/04/07
LOV relays may trip during grid faults
2B-04 Was De-energized on overcurrent
Calculated SC Exceed Equipment Ratings and Capabilities
Potential Protective Device Tripping for LOCA with degraded
voltage
2006 CDBI Violation - OPR153 did not address Seismic event
for identified condition
2006 CDBI Violation - Calculated SC exceeds equipment
ratings
09/21/06
2006 CDBI Violation - Calculated SC exceeds equipment
ratings
08/30/06
Cable Overload Protection for existing design not documented
Cable Overload Commitments
1P-29 TDAFP Outboard Bearing Reached Alert Alarm
06/15/09
1P-29 Turbine Outboard Bearing Temp High
09/15/09
1P-29 TDAFW Pump Outboard Turbine bearing Temp High
01/04/10
1P-29 Turbine Outboard Bearing High Temp Alarm
07/12/10
Engineering Evaluation for 1P-29 Temperature Alert
10/04/10
1TR-2000B PT 19 1P-29 Temperature High Alarm
01/10/11
Test Cables in CSR and 2PR-49 Usability Issue
05/17/11
Pump Secured Due to Outbrd Turb Bearing Temp > 250
Degrees F
06/16/11
Potential Overstresses Beams at EL. 26 of U2 Turbine
Building
7/13/11
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
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
DC Starter Verification & TOL Test for 2SMS-2019,
04/10/20
11
ICP 6.6 Service Water Instrumentation - Controlled
ICP 6.6 Service Water Instrumentation - Clean Side
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
Design Basis Document - 4160VAC System
5
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
MOV Capacity during LOOP/LOCA
0
MR 02-039* A/B
Aux Feed Water Pump 2-29 Recirculation Line Orifice
03/08/03
Unit 2 Main Steam and Feedwater Pipe Supports
0
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
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
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
Reactor Trip of Safety Injection
Reactor Trip Response
Rev 38
Loss of Reactor or Secondary Coolant
SI Termination
Post LOCA Cooldown and Depressurization
Faulted Steam Generator
Steam Generator Tube Rupture
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
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
Control Room Inaccessibility
6
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
Loop 2PT-4069 Functional Check
04/20/2011
RMP 9200-2
Station Battery D-06 Discharge Tests, Recovery and
Equalizing Charge
03/24/2009
125V Station Tech Spec Batteries Weekly Inspection
07/12/2011
125V Station Tech Spec Batteries Weekly Inspection
08/12/2011
125V Station Tech Spec Batteries Weekly Inspection
07/26/2011
D-06, Quarterly Station Battery Inspection
01/10/2011
D-06, Quarterly Station Battery Inspection
04/12/2011
D-06, Quarterly Station Battery Inspection per RMP 9046-1
06/21/2011
D-06, Annual Station Battery Inspection per RMP 9046-1
05/17/2010
D-06, Annual Station Battery Inspection per RMP 9046-1
05/04/2009
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
D-06 Modified Performance Test
05/04/2009
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
Alternating Current
Apparent Cause Evaluation
Agencywide Document Access Management System
Abnormal Operating Procedure
Action Request
American Institute of Steal Construction
American Society of Mechanical Engineers
Component Design Bases Inspection
CFR
Code of Federal Regulations
Condensate Storage Tank
Division of Reactor Safety
Emergency Operating Procedure
Extended Power Uprate
°F
Fahrenheit Degrees
Finding
GL
Generic Letter
IMC
Inspection Manual Chapter
IN
Information Notice
IR
Inspection Report
Inservice Testing
kV
Kilovolt
Loss of Coolant Accident
LONF
Loss of Normal Feedwater
Loss of Off-site Power
Motor Driven Auxiliary Feedwater
Motor-Operated Valve
Non-Cited Violation
Net Positive Suction Head
NRC
U.S. Nuclear Regulatory Commission
Operational Decision Making
Operation and Maintenance
Publicly Available Records System
psig
Pressure Per Square Inch Gage
Regulatory Issue Summary
Station Blackout
Significance Determination Process
Turbine Driven Auxiliary Feedwater
TS
Technical Specification
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.
Enclosure:
Inspection Report 05000266/2011009; 05000301/2011009
w/Attachment: Supplemental Information
cc w/encl:
Distribution via ListServ
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OFFICE
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NAME
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DATE
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10/17/11
OFFICIAL RECORD COPY