ML100060183
| ML100060183 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/06/2010 |
| From: | Ann Marie Stone NRC/RGN-III/DRS/EB2 |
| To: | O'Connor T Northern States Power Co |
| References | |
| IR-09-007 | |
| Download: ML100060183 (48) | |
See also: IR 05000263/2009007
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE ROAD, SUITE 210
LISLE, IL 60532-4352
January 6, 2010
Mr. Timothy J. OConnor
Monticello Nuclear Generating Plant
Northern States Power Company, Minnesota
2807 West County Road 75
Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT NRC COMPONENT DESIGN
BASES INSPECTION (CDBI) INSPECTION REPORT
Dear Mr. OConnor:
On December 4, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed a
component design bases inspection at your Monticello Nuclear Generating Plant. The
enclosed report documents the inspection findings, which were discussed on October 2, 2009,
and on December 4, 2009, with you and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your
license. The inspectors reviewed selected procedures and records, observed activities, and
interviewed personnel.
Based on the results of this inspection, five NRC-identified findings of very low safety-
significance were identified. The findings involved a violation 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 VI.A.1 of the NRC Enforcement Policy.
If you contest the subject or severity of these NCVs, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the 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 Monticello Nuclear Generating Plant. In addition, if you
disagree with the characterization of 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 Monticello Nuclear
Generating Plant. The information that you provide will be considered in accordance with
Inspection Manual Chapter 0305.
T. OConnor
-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 Agencywide Documents Access and Management System (ADAMS),
accessible from the NRC Web site 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 No. 50-263
License No. DPR-22
Enclosure:
Inspection Report 05000263/2009007
w/Attachment: Supplemental Information
cc w/encl:
Distribution via ListServ
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
50-263
License No:
Report No:
Licensee:
Northern States Power Company, Minnesota
Facility:
Monticello Nuclear Generating Plant
Location:
Monticello, MN
Dates:
August 31 through December 4, 2009
Inspectors:
A. Dunlop, Senior Reactor Engineer, Lead
N. Féliz Adorno, Reactor Engineer, Mechanical
C. Brown, Reactor Engineer, Operations
J. Bozga, Reactor Engineer, Mechanical
M. Munir, Reactor Engineer, Electrical
S. Kobylarz, Electrical Contractor
H. Campbell, Mechanical Contractor
Observers:
L. Jones, Reactor Engineer
S. Edmonds, Reactor Engineer
Approved by:
Ann Marie Stone, Chief
Engineering Branch 2
Division of Reactor Safety
Enclosure
SUMMARY OF FINDINGS
IR 05000263/2009007; 08/31/2009 - 12/04/2009; Monticello Nuclear Generating Plant;
Component Design Bases Inspection (CDBI) and Power Uprate
The inspection was a 3-week onsite baseline inspection that focused on the design of
components that are risk-significant and have low design margin; and power uprate. The
inspection was conducted by regional engineering inspectors and two consultants. Five Green
findings were identified by the inspectors. 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: Barrier Integrity
Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III,
Design Control, having very low safety-significance for the failure to incorporate the
actual physical configuration of the inboard main steam isolation valves (MSIVs) and the
correct pneumatic system pressure drop into the pneumatic pressure requirement
calculation for the inboard MSIVs. Specifically, the licensee failed to adjust the actuator
moving part weight to reflect that the actuator was offset by 45 degrees instead of being
mounted vertically and to correctly compute the system pressure drop. This finding was
entered into the licensees corrective action program and a preliminary calculation
performed by the licensee concluded that the valves were operable.
The finding was more than minor because it was associated with the Barrier Integrity
cornerstone attribute of structures, systems, components, and barrier performance, and
affected the cornerstone objective of providing reasonable assurance that physical
design barriers 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. (Section 1R21.3.b.(1))
Cornerstone: Mitigating Systems
Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III,
Design Control, having very low safety-significance for the failure to restore the
emergency service water (ESW) piping supports to their design specifications.
Specifically, although the licensee identified the existence of gaps between the ESW
piping supports and the baseplates, the licensee failed to recognize that this condition
did not meet seismic Category 1 design basis requirements. As a result, corrective
actions were not implemented. The licensee entered this issue into its corrective action
program and restored the supports to their design specifications.
Enclosure
1
The finding was more than minor because it was associated with the Mitigating Systems
cornerstone attribute of protection against external events and affected the cornerstone
objective of ensuring the availability of the ESW system, and ultimately the emergency
diesel generators (EDGs), to respond to initiating events to prevent undesirable
consequences. This finding is of very low safety-significance (Green) because the
design deficiency was confirmed not to result in loss of operability or functionality. This
finding has a cross-cutting aspect in the area of problem identification and resolution
because the licensee did not properly prioritize and evaluate an identified problem.
P.1(c). (Section 1R21.3.b.(2))
Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III,
Design Control, having very low safety-significance for the failure to adequately
evaluate circuit loads in determining design limits in electrical calculations. Specifically,
three examples were identified where the licensee: (1) failed to perform a calculation for
safety-related motor starters that included all control circuit loads in determining the
minimum voltage available at 120Vac starter coils, which was used to establish the coil
voltage test acceptance criteria; (2) failed to include thermal overload heater and starter
contact resistance when calculating the minimum voltage at 480Vac motor terminals;
and (3) failed to assure that the minimum voltage at the 120Vac solenoid operated
control valves was in conformance with vendor requirements. These issues were
entered into the licensees corrective action program to re-evaluate the voltage available,
and to test coils, as required, to verify the pick-up voltage.
The finding was more than minor because it was associated with the Mitigating Systems
cornerstone attribute of design control and affected the cornerstone objective of ensuring
the availability, reliability, and capability of safety-related equipment to respond to
initiating events to prevent undesirable consequences. This finding is of very low safety-
significance (Green) because the design deficiency was confirmed, with the exception of
ESW pump P-111B, not to result in loss of operability or functionality. Specifically, the
failure to assure adequate voltage was available at the solenoid valves coils; and to
perform periodic testing to assure the minimum voltage remained acceptable as the
components aged, did not result in an impact on current operability. With respect to the
ESW pump, it was determined that the pump would not have started under degraded
voltage condition as required such that the ESW pump was considered inoperable.
Based on a Phase III analysis, the failure of the pump to start under degraded voltage
conditions was determined to be very low safety-significance (Green). The inspectors
did not identify a cross-cutting aspect associated with this finding because this was a
legacy design issue, therefore was not reflective of current performance. (Section
1R21.3.b.(3))
Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI,
ATest Control,@ having very low safety-significance for the failure to have adequate
testing for safety-related equipment to monitor component degradation. Specifically, the
licensee failed to verify that the motor control center contactors would continue to pick-
up under degraded voltage conditions with less than the vendors required minimum
voltage. These issues were entered into the licensees corrective action program to test
the 13 contactors as soon as practicable and to revise the maintenance procedures to
incorporate the requirements for periodic testing of contactors.
Enclosure
2
The finding was more than minor because it was associated with the Mitigating Systems
cornerstone attribute of design control and affected the cornerstone objective of ensuring
the availability, reliability, and capability of safety-related equipment to respond to
initiating events to prevent undesirable consequences. This finding is of very low
safety-significance (Green) because the testing deficiency was confirmed, with the
exception of ESW pump P-111B, not to result in loss of operability or functionality.
Specifically, subsequent testing confirmed for nine contactors that the safety-related
starter coils would still function at the calculated degraded voltage values. Although
three of the contactors have not been tested, they were of a different size than the failed
contactor and there appeared to be reasonable assurance based on the successful tests
that these contactors also remained operable. With respect to the ESW pump, the failed
test confirmed that the motor starter contactor would not pickup under degraded voltage
conditions due to mechanical binding of the contactor arm such that the ESW pump was
considered inoperable. Based on a Phase III analysis, the failure of the pump to start
under degraded voltage conditions was determined to be very low safety-significance
(Green). 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. (Section 1R21.3.b.(4))
Cornerstone: Mitigating Systems and Barrier Integrity
Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III,
Design Control, having very low safety-significance for the failure of two pipe supports
to meet their design requirements. Specifically, the calculation for pipe support SR-526
failed to use the minimum yield strength in determination of the allowable bending stress
of the pipe support baseplate as required in the American Institute of Steel Construction
code. In addition, the calculation for pipe support PS-16 failed to use the design basis
concrete compressive strength in determination of the anchor bolt allowable as required
in the licensees design specification. This finding was entered into the licensees
corrective action program and a preliminary analysis performed by the licensee
concluded that the pipe supports were operable but nonconforming.
The performance deficiency for pipe support SR-526 example was more than minor
because it was associated with the Mitigating Systems cornerstone attribute of design
control and affected the cornerstone objective of ensuring the availability, reliability, and
capability of the safety-related residual heat removal and core spray pumps. This finding
is of very low safety-significance (Green) because the design deficiency was confirmed
not to result in loss of operability or functionality. The performance deficiency for pipe
support PS-16 example was more than minor because it was associated with the Barrier
Integrity cornerstone attribute of design control and affected the cornerstone objective of
providing reasonable assurance that physical design barriers 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.
(Section 4OA5.1.b.(1))
B.
Licensee-Identified Violations
No violations of significance were identified.
Enclosure
3
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 that 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
The inspectors selected risk-significant components and operator actions for review
using information contained in the licensees PRA and the Monticello Standardized Plant
Analysis Risk (SPAR) Model, Revision 3P. In general, the selection was based upon the
components and operator actions having a risk achievement worth of greater than 1.3
and/or a risk reduction worth greater than 1.005. The operator actions selected for
review included actions taken by operators both inside and outside of the control room
during postulated accident scenarios. In addition, the inspectors selected operating
experience issues associated with the selected components.
The inspectors performed a margin assessment and detailed review of the selected risk-
significant components to verify that the design bases have been correctly implemented
and maintained. This design margin assessment considered original design reductions
caused by design modification, or 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 action, 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.
This inspection constituted 30 samples as defined in Inspection Procedure 71111.21-05.
Enclosure
4
.3
Component Design
a.
Inspection Scope
The inspectors reviewed the Updated Safety Analysis Report (USAR), 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 (IEEE) 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 that 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 that 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,
system health reports, operating experience-related information, and licensee corrective
action program documents. Field walkdowns were conducted for all accessible
components to assess material condition and to verify that 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 20 components were reviewed:
Emergency Diesel Generator (EDG) (DG 11): The inspectors reviewed the EDG
loading calculation and vendor ratings for conformance with design basis load
requirements. The inspectors also reviewed EDG vendor de-rating requirements
for potential impact on design basis loading and operating procedures to
determine whether de-rating requirements were incorporated appropriately. In
addition, the inspectors reviewed surveillance testing to determine whether
design basis load requirements were demonstrated during periodic load testing
to satisfy TS.
Diesel Fuel Oil Transfer and Service Pumps (P-11 and P-77): The inspectors
reviewed the system hydraulic calculations including net positive suction head
(NPSH) and vortexing to ensure that the pumps were capable of providing
sufficient flow such that the day tanks remained filled during diesel operation.
The inspection also included a review of operating procedures related to these
functions. The inspectors reviewed vendor specifications and pump curves to
make sure that these parameters had been correctly translated into calculations,
as required. In addition, design change history, corrective actions, surveillance
results, and trending data were reviewed to assess potential component
degradation and impact on design margins.
Enclosure
5
EDG 11 Room Louvers/Fan (V-SF-10): The inspectors reviewed calculations
and operating procedures to ensure that the ventilation system was capable of
maintaining room temperatures within design limits. The review included the
control logic associated with the ventilation system to ensure it would function as
designed. Seismic Category I qualification of ventilation ductwork supports in
EDG building were also reviewed. The inspectors reviewed vendor
specifications to verify these parameters had been correctly translated into
calculations, as required. In addition, design change history, corrective actions,
surveillance results, and trending data were reviewed to assess potential
component degradation and impact on design margins.
Emergency Service Water (ESW) Pump (P-111A): The inspectors reviewed the
system hydraulic calculations including NPSH, vortexing, and waterhammer to
ensure the pump was capable of providing sufficient flow under accident
conditions. The review also included the control logic and bay water level
setpoint associated with the pump to ensure it would function as designed.
Operating procedures related to the pump were also reviewed. The inspectors
reviewed vendor specifications and pump curves to verify these parameters had
been correctly translated into calculations, as required. In addition, design
change history, corrective actions, surveillance results, and trending data were
reviewed to assess potential component degradation and impact on design
margins. The inspectors reviewed the operability evaluation for the piping and
pipe support analysis for ESW 1B and ESW 2B to verify there was a reasonable
assurance of operability. The motor sizing and pump brake horsepower
requirements and vendor ratings were reviewed for conformance with design
basis load conditions. The inspectors also reviewed design calculations to
determine the adequacy of voltage at motor terminals and motor starter
contactors during degraded voltage conditions and the adequacy of motor feeder
cable sizing. The motor and feeder cable coordination calculation was reviewed
to determine the adequacy of protection and coordination.
Main Steam Isolation Valve (MSIV) (AO-2-86D): The inspectors reviewed thrust
calculations associated with actuator thrust and pneumatic supply, and
environmental qualifications of the valve actuator to ensure the valve was
capable of functioning under design conditions. In addition, the inspectors
reviewed completed surveillances to ensure actual performance was acceptable
and vendor specifications to ensure parameters had been correctly translated
into calculations. Procedures related to the MSIVs were also reviewed. Design
change history, corrective actions, surveillance results, and trending data were
reviewed to assess potential component degradation, impact on design margin.
Residual Heat Removal (RHR) Pumps (P-202A/C): The inspectors reviewed the
pumps capability to meet design basis assumptions with respect to pump flow
and pressure. Calculations, drawings, procedures, tests, and other analyses
were reviewed to verify selected calculation inputs, requirements, and
methodologies were accurate and justified, and were consistently applied. The
inspectors reviewed completed surveillance tests to confirm the acceptance
criteria and test results demonstrated the capability of the pump to provide
required flow rates. Inservice Test (IST) results were reviewed to assess
potential component degradation and impact on design margins. In addition, the
licensee responses and actions to NRC Bulletin 88-04, Potential Safety-Related
Enclosure
6
Pump Loss, were reviewed to assess implementation of operating experience.
The calculation which developed the extent of containment overpressure
required to satisfy NPSH was also reviewed, in addition to emergency operating
procedures that directly incorporated this information. The inspectors reviewed
motor sizing and pump brake horsepower requirements and vendor ratings for
conformance with design basis load conditions. The inspectors also reviewed
design calculations to determine the adequacy of voltage at motor terminals
during degraded voltage conditions and the adequacy of feeder cable sizing.
The motor and feeder cable coordination calculation was reviewed to determine
the adequacy of protection and coordination.
RHR Pump Room Cooler (V-AC-4): The inspectors reviewed design calculations
associated with room heat loads and cooling to ensure the pump and associated
components remained within their temperature limits. Surveillance test results,
including verification of adequate air and water flows were also reviewed to
assess the capability of the room cooler to maintain temperature within
prescribed limits. In addition, two operability evaluations concerning room cooler
water flow test results and associated flow instrumentation were reviewed,
including the recently installed high accuracy flow instrumentation.
RHR Pump Miniflow Valve (CV-1994): Setpoint and scaling calculations were
reviewed to ensure that the vendor requirements per NRC Bulletin 88-04 for
minimum flow conditions were established. Maximum expected differential
pressure calculations, coupled with valve actuator performance requirements
were also reviewed to ensure the valve was capable of functioning under design
conditions. The inspectors reviewed the electrical schematic diagrams and
investigated the adequacy of voltage at the solenoids under design basis
accident conditions. The inspectors verified the adequacy of the feeder circuits
including the circuit breakers and its settings and cables. Finally, surveillance
test results also were reviewed to assess potential component degradation and
impact on design margins.
RHR B Loop Discharge to Torus Isolation Valve (MO-2007): The inspectors
reviewed motor-operated valve (MOV) calculations and analysis to ensure the
valve was capable of functioning under design conditions. These included
calculations for required thrust, maximum differential pressure, pressure locking,
seismic Category I qualification of pipe support SR-852, and valve weak link
analysis. The inspectors reviewed the electrical schematic diagrams and the
degraded voltage calculations for both the power and control circuits of the MOV.
The inspectors also verified the adequacy of the feeder circuits including the
circuit breakers and its settings, the power cables, and the thermal overload
relays. Diagnostic testing and IST surveillance results, including stroke time
testing, were reviewed to verify acceptance criteria were met and performance
degradation could be identified.
RHR B Loop Low Pressure Coolant Injection Outboard Injection Valve
(MO-2013): The inspectors reviewed MOV calculations and analysis to ensure
the valve was capable of functioning under design conditions. These included
calculations for required thrust, maximum differential pressure, seismic
Category I qualification of piping and pipe supports, and valve weak link analysis.
The inspectors reviewed the electrical schematic diagrams and the degraded
Enclosure
7
voltage calculations for both the power and control circuits of the MOV. The
inspectors also verified the adequacy of the feeder circuits including the circuit
breakers and its settings, the power cables, and the thermal overload relays.
Diagnostic testing and IST surveillance results, including stroke time and leak
rate testing, were reviewed to verify acceptance criteria were met and
performance degradation could be identified.
RHR Containment Spray Outboard Isolation Valve (MO-2020): The inspectors
reviewed MOV calculations and analysis to ensure the valve was capable of
functioning under design conditions. These included calculations for required
thrust, maximum differential pressure, seismic Category I qualification of pipe
support SR-680, and valve weak link analysis. The inspectors reviewed the
electrical schematic diagrams and the degraded voltage calculations for both the
power and control circuits of the MOV. The inspectors also verified the adequacy
of the feeder circuits including the circuit breakers and its settings, the power
cables, and the thermal overload relays. Diagnostic testing and IST surveillance
results, including stroke time and leak rate testing, were reviewed to verify
acceptance criteria were met and performance degradation could be identified.
In addition, a modification to provide operations the ability to throttle the valve
was reviewed.
Residual Heat Removal Service Water (RHRSW) Pumps (P-109A/C): The
inspectors reviewed system hydraulic calculations including those addressing
NPSH and vortex concerns related to the intake structure and associated
required water levels. Further, calculations and the adequacy of the differential
pressure setpoint across the RHR heat exchanger were reviewed to ensure the
service water side was at a higher pressure than the RHR side. The inspectors
reviewed pump installation and manufacturing details, vendor manuals,
specifications, and pump curves to make sure that these parameters had been
correctly translated into calculations, as required. Motor sizing and pump brake
horsepower requirements and vendor ratings were reviewed for conformance
with design basis load conditions. The inspectors also reviewed design
calculations to determine the adequacy of voltage at motor terminals during
degraded voltage conditions and the adequacy of feeder cable sizing. The motor
and feeder cable coordination calculation was reviewed to determine the
adequacy of protection and coordination. In addition, the inspectors reviewed
completed pump surveillances to ensure that actual performance was acceptable
and no significant degradation was taking place.
RHRSW Flow Control Valve (CV-1729): The inspectors reviewed maximum
expected differential pressures calculations, coupled with valve actuator
performance requirements to ensure the valve was capable of functioning under
design conditions. The calculation addressing the control feature of this valve, to
ensure a positive differential pressure of service water to RHR was also
reviewed. The inspectors reviewed the electrical schematic diagrams and
investigated the adequacy of voltage at the solenoids under design basis
accident conditions. The inspectors also verified the adequacy of the feeder
circuits including the circuit breakers and its settings and cables. Surveillance
and diagnostic test results were reviewed to ensure that actual performance was
acceptable and no significant degradation was taking place.
Enclosure
8
Enclosure
9
Reserve Auxiliary Transformer (1AR): The inspectors reviewed one-line
diagrams and vendor test results for impedance data to confirm that correct
transformer impedances were utilized in design analyses. The inspectors
confirmed the adequacy of the overcurrent relay setting calculation for design
basis loading and protective relay setting requirements. Surveillance testing for
the overcurrent relays was reviewed for issues that affect reliability and for
conformance with relay setting cards. The inspectors reviewed the modification
that installed the transformer for potential impact on the design basis.
345-4.16kV Low Voltage Auxiliary Transformer (2R): The inspectors reviewed
the design basis descriptions, equipment specifications, drawings, equipment
nameplate data, voltage drop calculations, and short circuit and load flow
calculations to evaluate the capability of transformer 2R to supply the voltage and
current requirements to station safeguard loads. Protective relay trip setting
calculations were reviewed to verify whether adequate protection coordination
margins were provided. The relay settings review included the transformer
overall differential currents and ground overcurrent relays. The inspectors
reviewed the results of completed transformer preventive maintenance and relay
calibrations to verify that the test results were satisfactory.
4.16 kV Switchgear (Bus 15): The inspectors reviewed load flow and short circuit
current calculations to determine the design basis for maximum load, interrupting
duty and bus bracing requirements, and the switchgear equipment vendor ratings
for conformance with design basis. The coordination/protection calculation for
the incoming line and feeder breaker relay settings was reviewed for design
basis loading and protective relay setting requirements. The inspectors reviewed
surveillance testing for the overcurrent relays for issues that affect reliability and
for conformance with relay setting cards. The inspectors also reviewed selected
breaker preventive maintenance for any recurring issues that affect reliability.
480V Load Center 103 (LC 103): The inspectors reviewed load flow and short
circuit current calculations to determine the design basis for maximum load,
interrupting duty and bus bracing requirements and the load center equipment
vendor ratings for conformance with design basis. The coordination/protection
calculation for the incoming line and feeder breaker relay settings was reviewed
for design basis loading and protective relay setting requirements. The
inspectors reviewed surveillance and preventive maintenance testing for the
breakers for issues that affect reliability and for conformance with protective
device trip requirements.
480V Motor Control Center (MCC 103A): The inspectors reviewed load flow and
short circuit current calculations to determine the design basis for maximum load,
interrupting duty and bus bracing requirements and the motor control center
equipment vendor ratings for conformance with design basis. The
coordination/protection calculation for the incoming line and feeder breaker relay
settings was reviewed for design basis loading and protective relay setting
requirements. The inspectors reviewed bus and motor starter surveillance and
preventive maintenance testing for issues that affect reliability.
125Vdc Station Battery/Battery Charger (D11/D10): The inspectors reviewed
calculations and analyses relating to battery sizing and capacity, hydrogen
generation, station blackout (SBO) coping, and battery room transient
temperature. The review was performed to ascertain the adequacy and
appropriateness of design assumptions, and to verify that the battery was
adequately sized to support the design basis required voltage requirements of
the 125Vdc safety-related loads under both design basis accident and SBO
conditions. The inspectors reviewed calculations relating to sizing and current
limit setting to ascertain the adequacy and appropriateness of design
assumptions, and to verify that the charger was adequately sized to support the
design basis duty cycle requirements of the 125Vdc safety-related loads and the
associated battery under both normal and design basis accident conditions. The
inspectors also reviewed a sampling of completed surveillance tests, service
tests, performance discharge tests, and modified performance tests. The review
of various discharge tests was to verify that the battery capacity was adequate to
support the design basis duty cycle requirements and to verify that the battery
capacity meets TS requirements. In addition, the test procedures were reviewed
to determine whether maintenance and testing activities for the battery charger
were in accordance with vendor=s recommendations. The inspectors also
witnessed a weekly battery surveillance test. Seismic Category I qualification of
the battery charger D10 anchorage was also reviewed.
125Vdc Distribution Panel (D11, D111): The inspectors reviewed 125Vdc short
circuit calculations and verified that the interrupting ratings of the fuses and the
molded-case circuit breakers were well above the calculated short circuit
currents. The 125Vdc voltage calculations were reviewed to determine if
adequate voltage would be available for the breaker open and close coils and
spring charging motors. The inspectors reviewed the motor control logic
diagrams and the 125Vdc voltage drop calculation to ensure adequate voltage
would be available for the control circuit components under all design basis
conditions. The inspectors also reviewed the 125Vdc short circuit and
coordination calculations to assure coordination between the motor feed breaker
open and close control circuit fuses and 125Vdc supply breakers and to verify the
interrupting ratings of the control circuit fuses and the 125Vdc control power feed
breaker.
b.
Findings
(1) Calculation Errors Associated With the Pneumatic Pressure Requirements for the
Inboard Main Steam Isolation Valves (MSIVs)
Introduction: A finding of very low safety-significance and associated Non-Cited
Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by
the inspectors for the failure to incorporate the actual physical configuration of the
inboard MSIVs and the correct pneumatic system pressure drop into the pneumatic
pressure requirement calculation for the inboard MSIVs.
Description: On September 2, 2009, the inspectors identified that the licensee failed to
adjust the weight of the actuator moving parts for the inboard MSIVs to reflect that the
actuator was offset by 45 degrees instead of being mounted vertically as assumed in
calculation CA-94-037, Calculation of Inboard MSlV Post-LOCA Closing Forces and
Enclosure
10
Pneumatic Pressure Requirements. This error affected the calculated minimum
alternate nitrogen system operating pressure required to close the inboard MSIVs.
The MSIVs were part of the primary containment with safety functions that included:
(1) providing primary containment isolation; (2) preventing core damage by limiting the
loss of coolant from the reactor vessel following a main steam line break outside the
primary containment; and (3) preventing excessive release of radioactivity to the
environs under assumed conditions of a primary steam line break outside of the primary
containment. The inboard MSIVs were wye pattern globe valves. The main disc was
attached to the lower end of the stem and moves in the valve guides at a 45 degree
angle from the inlet pipe. The valves were provided with springs that maintain the valves
in the closed position once they were shut by the valves air operator.
The normal pilot and safety grade pneumatic supplies for the inboard MSIVs was
provided by the alternate nitrogen system. The manifold and system pressures of the
alternate nitrogen system were monitored by pressure switches that give control room
annunciation on low pressure. The nominal setpoint for the pressure switches was
91 psig. However, the pressure switch has an instrument uncertainty of 3.7 psig, such
that the alarm potentially would not actuate until actual pressure was 87.3 psig.
The calculated minimum pneumatic pressure required to close the inboard MSIVs was
87 psig assuming that the valves were installed vertically. However, since the valves
were at a 45 degree angle, the closing force provided by the weight of the actuator
moving parts would be less than that assumed in the calculation. In addition, the
alternate nitrogen system pressure drop was incorrectly determined because of a
computational error. When the calculation incorporated the actual configuration of the
valves and the corrected pressure drop, the minimum required pneumatic pressure
increased to 88.84 psig, which was greater than the potential minimum pressure of 87.3
psig allowed by plant procedures.
The licensee evaluated the pneumatic pressure requirements for the inboard MSIVs and
determined that the valves were operable by removing conservatisms in the affected
calculation. The inspectors performed an independent review of the evaluation and had
no further concerns. The licensee initiated action requests (ARs) 01196394 and
01198495 to revise the affected calculation.
Analysis: The inspectors determined that the failure to incorporate the actual physical
configuration of the inboard MSIVs and the correct pneumatic system pressure drop into
the pneumatic pressure requirement calculation of the inboard MSIVs was a
performance deficiency.
The performance deficiency was determined to be more than minor because it was
associated with the Barrier Integrity cornerstone attribute of structures, systems,
components and barrier performance, and affected the cornerstone objective of
providing reasonable assurance that physical design barriers protect the public from
radionuclide releases caused by accidents or events. Specifically, the engineering
calculation error was significant enough to require the calculation to be re-performed to
assure that the minimum operating pressure of the alternate nitrogen supply was
adequate to support the MSIVs pneumatic requirements.
Enclosure
11
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 3b for the Barrier
Integrity cornerstone. The finding screened as of very low safety-significance (Green)
because it was a design deficiency of the physical integrity of the reactor containment
that: (1) did not affect the barrier function of the control room against smoke or a toxic
atmosphere; (2) did not represent an actual open pathway in the physical integrity of
reactor containment; and (3) did not involve an actual reduction in function of hydrogen
igniters in the reactor containment. Specifically, the last alternate nitrogen system
surveillance demonstrated that the pressure was higher than the corrected minimum
required pressure. In addition, the licensee showed that the calculation had sufficient
conservatisms to account for the calculation errors. The inspectors performed an
independent review of this evaluation and had no further concerns.
The inspectors determined there was no cross-cutting aspect associated with this finding
because this was a legacy design issue and therefore was not reflective of current
performance.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,
in part, that measures shall be established to assure that applicable regulatory
requirements and the design basis are correctly translated into specifications, drawings,
procedures, and instructions.
Contrary to the above, as of September 4, 2009, the licensee failed to correctly translate
applicable design basis into specifications. Specifically, calculation CA-94-037 failed to
adjust the closing force provided by the weight of the actuator moving parts to reflect that
the actuator was at a 45 degree angle and to correctly compute the pneumatic system
pressure drop. Because this violation was of very low safety-significance and it was
entered into the licensees corrective action program (ARs 01196394 and 01198495),
this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC
Enforcement Policy (NCV 05000263/2009007-01).
(2) Emergency Service Water (ESW) Piping Supports Did Not Meet Seismic Category 1
Design Basis Requirements
Introduction: A finding of very low safety-significance and associated Non-Cited
Violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was
identified by the inspectors for the failure to restore the ESW piping supports to their
design specifications.
Description: On September 9, 2009, the inspectors identified that the licensee failed to
recognize that the ESW discharge pipe supports did not meet their seismic Category 1
design basis requirements and that, as a result, the issue was not properly evaluated
such that corrective actions were not implemented.
During a plant walkdown, the inspectors noted that ESW piping stanchion supports
SWH-42 and SWH-47 were not in contact with their baseplates. The supports were
located on the discharge side of their respective ESW pumps near the boundary of
safety and non-safety-related piping. The gaps between the supports and the floor
baseplates were about 1/4 inch. These supports were classified seismic Category 1
because their function was to protect the safety-related piping that provides the required
Enclosure
12
cooling water to the EDGs. The EDGs were relied upon to safely shut down the reactor
upon the loss of all outside power simultaneous with a design basis accident or event.
Upon further review, it was discovered that the licensee had previously identified this
condition on at least three separate occasions.
On February 3, 2003, AR 00633248 documented the discovery of gaps at supports
SWH-42 and SWH-47. The piping was evaluated and found to be operable
assuming that the supports were not functional. In addition, the licensee repaired the
supports by removing the gaps to restore the piping and associated supports to
within ASME Code allowables. The AR was classified Level B, which meant that the
AR documented a condition that typically results in moderate impact to the plant.
On April 28, 2008, ARs 01135806 and 01135808 documented the discovery of the
gaps at the same supports. On this occasion, the licensee referred to the evaluation
conducted in 2003 and determined that no corrective actions were necessary
because the piping had been determined to be operable. Both ARs were classified
Level D, which meant that the ARs were considered to be associated with a
condition not adverse to quality.
On February 26, 2009, AR 01170994 documented the rediscovery of the gaps and
noted that the condition was previously evaluated in 2003 and 2008. No corrective
actions were implemented. The AR was classified Level C, which meant that the AR
documented a condition that typically results in minor impact to the plant.
It was also noted that in 2006, two check valves were removed and an air operated
valve was replaced with a manual valve on each ESW lines under Engineering Change
(EC) 768, Service Water Check Valves Replacement, resulting in a weight reduction on
the pipe. This could have potentially resulted in the condition identified in 2008;
however, this change was not evaluated as part of the 2008 or 2009 ARs.
The licensee initiated AR 01196345 to address the condition of the supports and
AR 01196433 to document that the condition adverse to quality was not corrected when
it was identified in 2008 and 2009. The piping was evaluated by the licensee and found
to be operable assuming that the supports were not functional. The inspectors
performed an independent review of this evaluation and had no further concerns. The
licensee repaired the supports by installing shims to fill in the gaps through work
requests 49141 and 49142.
Analysis: The inspectors determined that failure to restore the ESW piping supports to
their design specifications was a performance deficiency.
The performance deficiency was determined to be more than minor because it was
associated with the Mitigating Systems cornerstone attribute of protection against
external events and affected the cornerstone objective of ensuring the availability of the
ESW system, and ultimately the EDGs, to respond to initiating events to prevent
undesirable consequences. Specifically, the gaps between the ESW piping supports
and the baseplates created a condition were the supports did not meet their seismic
Category 1 design basis requirements. Furthermore, the finding impacted the availability
of the EDGs because they cannot operate for a long period of time without cooling
provided by ESW.
Enclosure
13
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 3b for the Mitigating System
cornerstone. The finding screened as very low safety-significance (Green) because the
finding was a design deficiency confirmed not to result in loss of operability or
functionality. Specifically, the licensee performed an operability evaluation that
concluded the ESW piping remained operable assuming the affected supports were not
functional. The inspectors performed an independent review of the operability
evaluation and did not have further concerns.
This finding has a cross-cutting aspect in the area of problem identification and
resolution, because the licensee did not thoroughly evaluate the issue such that the
corrective actions were inadequate. Specifically, the licensee failed to implement
corrective actions during the subsequent occasions that the condition was identified
because the problem was not properly prioritized and evaluated. The ARs initiated
subsequent to the condition being identified in 2003 were closed without corrective
actions under the premise that the condition was determined to be acceptable in 2003.
However, while the evaluation conducted in 2003 determined that the supports were
operable, it also determined that the supports needed to be restored to the original
design specifications. P.1(c)
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action,
requires, in part, that measures shall be established to assure that conditions adverse to
quality, such as failures, malfunctions, deficiencies, deviations, defective material and
equipment, and non-conformances are promptly identified and corrected.
Contrary to the above, in April 28, 2008 and February 26, 2009, the licensee failed to
promptly correct a condition adverse to quality regarding the ESW piping supports.
Specifically, although the licensee identified the gaps between the ESW piping supports
and the baseplates, it failed to recognize that this condition did not meet the seismic
Category 1 design basis requirements. Because the licensee failed to recognize the
qualification requirement, the condition was evaluated incorrectly and determined to be
acceptable such that corrective actions were not implemented. Because this violation
was of very low safety-significance and it was entered into the licensees corrective
action program (ARs 01196345 and 01196433), this violation is being treated as an
NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000263/2009007-02).
(3) Failure to Adequately Evaluate Minimum Voltage Available at Safety-Related Electrical
Components
Introduction: A finding of very low safety-significance and associated Non-Cited
Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by
the inspectors for the failure to adequately evaluate circuit loads in determining design
limits in electrical calculations. Specifically, the inspectors identified three examples
associated with the following: (1) the licensees failure to perform a calculation for
safety-related motor starters that included all control circuit loads in determining the
minimum voltage available at 120Vac starter coils, which was used to establish the coil
voltage test acceptance criteria; (2) the licensees failure to include thermal overload
(TOL) heater and starter contact resistance when calculating the minimum voltage at
480Vac motor terminals; and (3) the licensees failure to assure that the minimum
Enclosure
14
voltage at the 120Vac solenoid operated control valves was in conformance with vendor
requirements.
Description: The inspectors identified three examples where the licensee failed to
adequately evaluate circuit loads in determining design limits in electrical calculations.
These examples are as follows:
Failure to Evaluate Circuit Loads in Determining the Minimum Voltage Available At
Safety-Related Motor Starter Contactors
The inspectors reviewed the licensee evaluation on IN 2007-09, Equipment
Operability Under Degraded Voltage Conditions, and found that the licensee had not
adequately reviewed the notice for the issue on the adequacy of voltage at motor
starter coils during degraded voltage conditions. The licensee initiated AR 01196447
to perform the required review. The inspectors reviewed calculations CA-92-250,
Project 92Q600 Cable Sizing Calculation, and CA-94-094, MCC Starter Coil
Pickup Voltages and Maximum Cable Lengths, for the motor starter coil voltage for
ESW pumps P-111A and P-111B respectively. The inspectors found that calculation
CA-94-094 determined 96.32 volts were available at the P-111B motor starter
contactor coil and subsequently used 96 volts as the coil test acceptance criteria on
drawing NE-36347-1A, 480V MCC Schedules. A starter contactor coil test was
used in lieu of meeting the vendor criteria for minimum pickup voltage of 102 volts,
which was 85 percent of rated voltage. However, the inspectors found that the
calculation failed to consider the load on the control power transformer (CPT) due to
timing relay burden, indicating light load, and contact resistance in determining the
limiting voltage available at the motor starter.
The licensee performed a hand calculation during the inspection, which determined
that less than 96 volts was available when all the circuit resistance elements were
included in the voltage analysis and established 94 volts as the new acceptance
criteria for testing the coil. The contactor coil was previously tested in 1994 to pickup
at 84 volts, which provided the basis for immediate operability on AR 01200723. As
additional corrective actions, the licensee planned to test all the affected coils on
drawing NE-36347-1A. On October 15, 2009, the test for the starter contactor coil
for P-111B determined that the voltage needed for the coil to pickup was 104 volts,
not the 94 volts required by the revised calculation. The licensee subsequently
replaced the starter with a spare unit that was functionally tested at 80 volts. The
licensee sent the failed starter to a test laboratory to determine the cause of failure
and to initiate an evaluation of extent of condition.
The test laboratory determined that the cause of the failure was mechanical binding
of the starter right side contact arm and plunger binding due to abrasion and
gouging, which then required a higher voltage for the coil to function. The test lab
was not able to determine the root cause of the abrasion and binding that lead to the
failure. Since this pump was operated periodically, the normal available voltage to
the starter contactor coil was sufficient to overcome the binding effect of the contact
arm such that the pump started as required. As part of the extent of condition for this
issue, the licensee established a motor starter action plan to test 12 additional
safety-related starters. In addition, motor starter testing methodology and preventive
maintenance tasks were under review such that this issue could be identified prior to
failure.
Enclosure
15
Failure to Fully Evaluate Power Circuit Resistance in Determining Minimum Voltage
Available at Safety-Related Motor Terminals
The inspectors reviewed calculation CA-06-104, 480V MCC to Motor Terminal
Voltage Drop, and identified that the calculation did not consider TOL heater
resistance or starter contact voltage drop in determining the voltage available at the
motor terminals. For ESW Pump P-111A, this reduced the available voltage 0.2
volts, which reduced the voltage margin available to the minimum motor voltage
(414 volts) by 12 percent. The inspectors found that the voltage available was
approximately 0.45 percent more than the minimum voltage required by the motor
based on the motor rating and the pump load requirement. The licensee initiated
AR 01197431, which determined that although the available margin had been
reduced, the operability of the pump was not affected by the calculation error.
Failure to Assure Adequate Minimum Voltage at Safety-Related 120Vac Solenoid
Operated Control Valves
Section 8.10 of the USAR stated that the 120Vac instrument buses were maintained
between 108Vac and 132Vac (120 +/- 10 percent). Per MWI-3-M-2.01, AC Electrical
Load Study, 120Vac rated loads, including the solenoid operated control valves,
would have adequate voltage if the AC instrument buses were maintained within this
voltage range. The vendor specified a voltage range of 102V - 132V for the solenoid
operated control valves to operate properly. The published minimum operating
voltage, as well as the minimum voltage used for equipment qualification testing for
the solenoid valves (SV) was 102Vac. This value was identified in the ASCO
Nuclear Catalog-Nuclear Products Qualified to IEEE Specifications, in ASCO
Solenoid Catalog No. 31, and was used as the minimum test voltage for the ASCO
Test Report AQS-21678/TR.
The inspectors requested the voltage drop analysis for the SVs, however, the
licensee was unable to locate an analysis for the SVs. The licensee performed an
informal analysis for SV1729 and SV1994, which were in the inspectors scope, and
as part of an extent of condition, SV1728, SV1995, and SV1996. Based on this
analysis, the licensee determined that the voltages at SV1728 and SV1995 were
below the vendor specified minimum rating. As a result of the negative margin
identified for SV1728 and SV1995, the licensee tested three ASCO SVs, two of
which were the same models as the existing solenoids, to determine their minimum
pull-in and dropout voltages at various differential pressure values across the valve.
These tests concluded that the SVs were capable of energizing at voltages lower
than the specified 102Vac.
The licensee entered this issue in their corrective action program as AR 01199936 to
complete a formal voltage drop calculation for 120Vac supplies to safety-related
equipment and to restore voltage margin as a long-term solution. The licensee
documented the details of the testing for the sampled SVs in operability
recommendation OPR 199936-1. Based on the results of the tests, the licensee
determined that there was reasonable assurance that these valves would perform
their safety functions as designed and therefore the OPR classified the solenoids as
operable but nonconforming.
Enclosure
16
Analysis: The inspectors determined that the licensees failure to ensure adequate
voltage was available to energize the starter coils for 480Vac safety-related equipment
and the 120Vac solenoid valve coils was a performance deficiency because the
operability of safety-related equipment could not be assured and could have resulted in
a loss of function during a design basis accident concurrent with a degraded voltage
condition.
The performance deficiency was determined to be more than minor because it was
associated with the Mitigating Systems cornerstone attribute of design control and
affected the cornerstone objective of ensuring the availability, reliability and capability of
safety-related equipment to respond to initiating events to prevent undesirable
consequences. Specifically, the failure to ensure adequate voltage was available to
energize safety-related starter coils to supply 480Vac power to motors and 120Vac to
power SV coils would have affected the availability of their respective equipment to
respond to initiating events.
The inspectors evaluated the finding, with the exception of ESW pump P-111B, using
IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for
At-Power Situations, SDP Phase 1 screening. The finding screened as very low safety-
significance (Green) because the finding was a design deficiency confirmed not to result
in loss of operability or functionality. Specifically, the failure to assure adequate voltage
was available at the starter coils and SV coils; and to perform periodic testing to assure
the minimum voltage remained acceptable as the components aged did not result in an
impact on current operability.
The inspectors evaluated the finding of ESW pump P-111B using IMC 0609,
Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power
Situations, SDP Phase 1 screening. In accordance with Table 3b, "SDP Phase 1
Screening Worksheet for Initiating Events, Mitigating Systems, and Barriers
Cornerstones," the finding affected the safety of an operating reactor, specifically, the
Mitigating Systems Cornerstone. In accordance with Table 4a, "Characterization
Worksheet for Initiating Events, Mitigating Systems, and Barriers Cornerstones," the
finding represented a potential loss of a safety function of a single train of emergency AC
power for greater than its TS allowed outage time. The inspectors contacted a Region III
Senior Reactor Analyst (SRA) to perform an SDP review of the finding.
In accordance with IMC 0609, Appendix A, the SRA determined that the Phase 2 SDP
pre-solved tables showed that having one EDG service water pump inoperable for an
entire year was a Yellow finding. The SRA determined this result to be conservative
since P-111B was not completely failed, but shown to be nonfunctional only during
certain degraded voltage conditions. Further, the voltage of interest at the P-111B
starter coil was relevant when 12 EDG was supplying its associated safety-related 4160
Vac bus (Bus 16) and also during a loss of coolant accident (LOCA) with offsite power
available (with EDG 12 operating but not supplying Bus 16) and degraded voltage
conditions prior to a loss of off-site power LOOP occurring. The SRA performed an SDP
Phase 3 analysis to further characterize the significance of the finding.
The SRA performed a Phase 3 analysis using the NRC Risk Assessment
Standardization Project Handbook and the SPAR Model for Monticello, Revision 3P,
Level 1, Change 3.45, dated August 2008. Assuming failure to start of P-111B for a one
year exposure, the delta-CDF was less than 1E-7. The dominant cut-sets involved
Enclosure
17
station blackout events with failure of emergency power sources and failure to recover
either offsite or emergency power. The SRA contacted the licensee to review their risk
analysis.
The licensee reviewed recent emergency core cooling system (ECCS) test records,
which showed that P-111B started successfully under nearly identical voltage conditions
to those that would be expected during a loss of offsite power transient with an ECCS
initiation signal present. The licensee also performed a quantitative analysis assuming
a failure to start probability for P-111B of ten times its nominal value. The resultant
delta-CDF was about 3.6 E-7/yr. Nearly all of the increased risk was associated with
internal flooding scenarios where the flood initiates at or propagates to the lower
switchgear, which in turn causes a loss of offsite power event.
Given the low numerical risk results for the core damage sequences, the SRA concluded
that the risk associated with this performance deficiency was very low safety-significance
(Green).
Based on the results of the SDP Phase I and Phase III screenings, these three
examples for this finding screened as very low safety-significance (Green). The
inspectors determined there was no cross-cutting aspect associated with this finding
because the three examples were legacy design issues and therefore was not reflective
of current performance.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control requires,
in part, that design control measures provide for verifying or checking the adequacy of
design, such as by the performance of design reviews, by the use of alternate or
simplified calculational methods, or by the performance of suitable testing program.
Contrary to this requirement, prior to September 30, 2009, for P-111B and
September 28, 2009, for the solenoid valve coils, the licensees design control measures
failed to verify the adequacy of minimum design voltage for the safety-related coils.
Specifically, the licensee failed to adequately establish the degraded voltage value for
safety-related starters and solenoid valve coils. In addition, periodic testing at degraded
voltage, which was lower than the vendors rated minimum voltage, was not performed
to verify that the coils remained acceptable as the components aged. However,
because this violation was of very low safety-significance and because the issue was
entered into the licensees corrective action program (AR 01200723, AR 01197431, and
AR 01199936), this violation is being treated as an NCV consistent with Section VI.A.1
of the NRC Enforcement Policy. (NCV 05000263/2009007-03)
(4) Inadequate Testing for Motor Control Center (MCC) Contactors
Introduction: A finding of very low safety-significance and associated Non-Cited
Violation of 10 CFR Part 50, Appendix B, Criterion XI, ATest Control,@ was identified by
the inspectors for the failure to have adequate testing for safety-related equipment to
monitor component degradation. Specifically, the licensee failed to verify that the MCC
contactors would continue to pick-up under degraded voltage conditions with less than
the vendors required minimum voltage.
Description: The licensees AC load study identified that the lowest voltage at the MCCs
under degraded voltage conditions was 426V. In order for MOVs and other components
Enclosure
18
to operate, the MCC contactors needed to be verified to pick-up at the MCC degraded
voltage. Calculation CA-94-094, MCC Starter Coil Pick-Up Voltages and Maximum
Cable Lengths, determined the maximum cable length for each size of control cable to
ensure reliable circuit operation of contactors under degraded voltage conditions. This
calculation was performed to ascertain the maximum cable lengths with both 75 percent
and 85 percent contactor pick-up voltage for various combinations of National Electrical
Manufacturers Association (NEMA) starter sizes and CPT sizes at degraded voltage.
The calculation determined there were 13 contactors that had less than 85 percent
pick-up voltage, although only MO-2013 was included in the inspectors scope.
MO-2013 contactor was shown to have 84.52 percent pick-up voltage. The calculation,
however, did not account for valves MO-2007 and MO-2020, which were also in the
inspectors scope.
The licensee performed an informal evaluation for MO-2007 and MO-2020 and
determined that the cable lengths for both MOVs contactors were bounded by the
maximum cable lengths determined for a NEMA size 1 starter with a 75 volt amp CPT.
The inspectors reviewed this evaluation and determined that the assessment for these
two MOVs was acceptable. This evaluation also identified that there were interposing
relays installed in the MO-2013 control circuit to facilitate the pick-up of the contactor,
which were not accounted for in the calculation CA-94-094.
The licensee provided calculation CA-92-223, Analysis for Mod 92Q520 Control Circuit
Voltage Drop, to address the interposing relay issue. This calculation determined that
the starter coil for MO-2013 would have 83 percent of the rated voltage with the
interposing relays in the circuit, which was less than the 84.52 percent identified in
calculation CA-94-094. This calculation also stated a coil pick-up voltage acceptance
criterion of 75 percent instead of 85 percent as indicated in calculation CA-94-094. The
inspectors were concerned about the discrepant acceptance criteria used for the
contactor pick-up voltage. In response, the licensee stated that during the evaluation of
IN 94-050, Failure of GE Contactors to Pull in at the Required Voltage, it was
confirmed by General Electric (GE) that the contactor coils were rated for 85 percent and
not the 75 percent pick-up voltage indicated in calculation CA-92-223. The licensee
initiated AR 01200487 to supersede calculation CA-92-223 with calculation CA-94-094.
The inspectors expressed concern with the 13 GE contactors (associated with an MOV,
fan, or pump) having less than 85 percent pick-up voltage and prompted the licensee for
justifying operability under degraded voltage conditions. The licensee stated tests were
conducted in 1976, 1977, 1994, and 2000, on a few contactor samples. The testing
indicated that the contactors were able to pick up at less than 85 percent voltage.
However, the licensee did not have a periodic testing program in place to ensure that
age related degradation did not have an adverse impact on the contactors. The licensee
provided Procedure 4847-PM, GE 7700 Line Motor Control Center Maintenance
Procedure, which indicated testing would only be performed on a starter when it was
replaced. As such, the licensee was relying on the old test data to justify the continued
operability of these contactors.
The licensee entered this issue into their corrective action program and initiated
ARs 01200539 and 01200540 to revise the maintenance procedures to incorporate the
requirements for periodic testing of contactors. In addition, the licensee initiated work
orders to test the 13 contactors as soon as practicable. As of October 15, 2009, the
licensee had tested 10 of the contactors, which verified for 9 out the 10 that the
Enclosure
19
contactors would still pick-up at the calculated degraded voltage condition. However, as
discussed in Section 1R21.3.b.(3) of this report, the starter motor contactor associated
with ESW Pump P-111B failed during this testing. Additional corrective action and
extent of condition for this issue was discussed in Section 1R21.3.b.(3) of this report.
Analysis: The inspectors determined that the failure to have an adequate testing
program for the MCC contactors was a performance deficiency, because the failure of
the contactors to pick up could have resulted in a loss of function during design basis
accident conditions. Specifically, the licensee had determined from calculations that the
voltage at the 13 contactors would be less than the vendor required 85 percent pick-up
voltage and had not implemented any periodic testing to ensure the contactors would not
degrade over time. Failure of the contactors to pick-up would prevent the associate
MOV, fan, or pump to operate as required under a degraded voltage condition.
The performance deficiency was determined to be more than minor because it was
associated with the Mitigating Systems cornerstone attribute of design control and
affected the cornerstone objective of ensuring the availability, reliability and capability of
safety-related equipment to respond to initiating events to prevent undesirable
consequences. Specifically, the failure to verify that safety-related starter coils were not
degrading over time thorough periodic testing such that they would continue to function
at a lower calculated voltage. The failure of the starter coil to energize would prevent the
associate MOV, fan, or pump to operate as required under a degraded voltage condition.
The inspectors determined the finding, with the exception of ESW pump P-111B, 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 3b for the Mitigating System cornerstone. The finding screened as very
low safety-significance (Green) because the finding was a design deficiency confirmed
not to result in loss of operability or functionality. Specifically, the licensee confirmed
for 9 of the 13 contactors that they would function at a lower voltage than required by the
vendor. The inspectors reviewed the results of the tests and verified that the tested
contactors were able to pick up at less than 85 percent voltage. The inspectors
determined that the schedule submitted by the licensee for testing the remaining
contactors was reasonable. Based on these test results, no apparent common cause to
the failed starter motor, and initial contactor testing performed in 1976, 1977, 1994, and
2000, there was reasonable assurance of operability for the remaining untested
The inspectors Phase III screening of the ESW pump P-111B finding was discussed in
Section 1R21.3.b.(3) of this report, which concluded that the risk associated with this
performance deficiency was very low safety-significance (Green).
Based on the results of the SDP Phase I and Phase III screenings, this finding screened
as very low safety-significance (Green). The inspectors determined there was no
cross-cutting aspect associated with this finding because this was a legacy design issue
and therefore was not reflective of current performance.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XI, ATest Control,@ requires,
in part, that a test program shall be established to assure that all testing required to
demonstrate that structures, systems, and components will perform satisfactorily in
service is identified and performed in accordance with written test procedures, which
Enclosure
20
incorporate the requirements and acceptance limits contained in applicable design
documents.
Contrary to the above, as of October 2, 2009, the licensee failed to establish a periodic
testing program to ensure that the MCC contactors would be energized to perform its
safety-function under design basis accident conditions. Specifically, the licensee failed
to require periodic testing of contactors to ensure that age related degradation did not
have any adverse impact on contactors pick-up voltage. However, because this
violation was of very low safety-significance and because the issue was entered into the
licensees corrective action program (AR 01200539 and 01200540), this violation is
being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy.
(5) EDG Fuel Oil Supply System Design Does Not Meet Single Failure Criteria
Introduction: The inspectors identified an unresolved item (URI) regarding the EDG fuel
oil supply system. Specifically, the diesel fuel oil supply system design does not meet
the single failure criteria.
Description: The diesel fuel oil supply system consisted of two pumps, the fuel oil
transfer pump and the fuel oil service pump. However, only the fuel oil transfer pump
was considered safety-related. Nonetheless, TS Basis, Section 3.8.1, stated that both
pumps must be operable in order for both EDGs to be operable.
The original Monticello Safety Evaluation Report (SER), dated March 20, 1970, stated
that the EDGs were separate and independent with respect to fuel supplies. It also
concluded that the onsite emergency electric power system was acceptable since no
single failure should prevent power from being supplied to the engineering safety
features from onsite sources. In addition, the original Safety Analysis Report stated that
auxiliaries required to ensure continuous operation of the EDGs shall be supplied from
the essential buses or control power transformers associated with the EDGs. However,
the fuel oil service pump was powered by a non-essential motor control center that
would be load shed on an essential bus transfer load shed signal. In fact, EDG 11 does
not supply either pump without manual operator action. The fuel oil transfer pump was
normally powered by EDG 12.
In addition, the USAR, Section 10.3.1.5, stated that the safe-shutdown analysis
performed to comply with 10 CFR Part 50, Appendix R, Section G, revealed three fire
areas that required repair of equipment or use of alternate fuel oil pumping methods and
procedures for diesel oil pumping capability to the EDG day tanks to achieve cold
shutdown.
Further, the TS Basis, Section 3.8.1, described the diesel fuel oil supply piping as
redundant. However, the inspectors confirmed that the actual configuration was not
redundant in that both pumps discharge into common piping.
The inspectors also questioned if the service pump was tested to supply an adequate
flow. Specifically, TS Surveillance Requirement 3.8.1.5 verified, in part, that the fuel oil
transfer system operates to transfer fuel oil from the T-44 storage tank to the day tanks.
The TS Basis, Section 3.8.1, stated that this surveillance provides assurance that the
required fuel oil transfer pumps are operable. Operable was defined in TS 1.1, which
Enclosure
21
stated, in part, when the component is capable of performing its specified safety
function. The TS Basis, Section 3.8.1, also stated that the fuel oil transfer pump and the
fuel oil service pump are individually capable of maintaining the level in the day tank
when both EDGs are operating at full load. However, the inspectors noted that the
licensee only ensured that the fuel oil service pump transferred fuel and maintained level
in just one EDG day tank.
This issue is unresolved pending further NRC review of the licensing basis for the diesel
fuel oil transfer system and determination of NRC courses of action for resolution of the
issues. (URI 05000263/2009007-05).
(6) Inadequate Tornado Missile Protection for the EDG System Components
Introduction: The inspectors identified an unresolved item (URI) regarding the design
and licensing basis for the standby diesel generator (EDG) building ventilation system
and whether the ventilating system had to be protected from the effects of a design basis
tornado.
Description: During a walkdown of the EDG building, the inspectors noted that
temperature control dampers in the ventilation system were mounted flush with the
outside walls of the building with only a metal grating serving as a barrier for tornado
missiles. The inspectors questioned the adequacy of the design and received the
licensee position that the EDG building was a Class 1 building designed to protect the
EDG and the fuel oil day tank from tornado missiles. However, the licensee stated that
the temperature control dampers and ventilation fans (V-SF-9 and - 10) were not
designed to be protected from a tornado missile. The licensee stated that the designer
and the licensee intended on protecting Class 1 equipment required to assure safe-
shutdown of the reactor from a missile event based on the credibility of the missile. The
licensee agreed that the ventilation fans were necessary for the EDGs to be operable
and that the EDGs could only run for a short period of time without the fans running.
However, the licensee maintained that the EDGs could perform their safe-shutdown
function without the ventilation fans. The inspectors disagreed as the EDGs could not
run long enough to shut the reactor down, remove latent and decay heat, and maintain
the reactor in a cold shutdown condition following the loss of the EDG ventilation
systems. The licensee maintained that a single missile could not take out both fans as a
full height reinforced concrete wall separated the two ventilation systems.
The inspectors reviewed the original design submittal (NSP-1 dated October 17, 1969),
and found that the Class 1 equipment included the Standby Diesel Generator System,
and the Emergency Buses and other electrical gear to and including power equipment
required for safe-shutdown. The ventilation systems were not included in the list of
Class 1 equipment or in the list of Class 2 equipment. However, the licensee believed
the ventilation systems were considered Class 2 based on the last entry in the Class 2
list: All Other Piping and Equipment not listed under Class 1. After further research,
the licensee stated that there was no specific statement regarding the EDG building
ventilation in any licensing document.
The inspectors identified the following statements in various licensing documents.
Section 2.2.4, Standby Diesel Generating Building, of the USAR stated, The
principal function of this building is to provide a safe enclosure and protection for the
Enclosure
22
standby diesel generators and portions of the power distribution systems enclosed
therein.
Section 2.2.4.1, Structure Description, stated, in part, A north-south [sic] interior
wall of reinforced concrete extends the full height of the structure providing physical
separation of the diesel generator systems.
The original Monticello SER, Section 3.1.2, Meteorology, stated, in part, that the
facility structures and systems, which are necessary for a safe-shutdown of the
reactor are designed to withstand the effects of wind loadings and potential missiles
resulting from a tornado.
In the current Revision 25 of the USAR the inspectors noted the following:
a)
Section 8.4.1.1, Design Basis, stated that two independent EDGs provide
redundant standby power sources.
b)
Section 8.4.1.1.b stated, The EDG sets shall be complete package units with
all auxiliaries necessary to make them self-sufficient power sources capable of
automatic start at any time and capable of continued operation at rated full load
voltage and frequency until either manually or automatically stopped.
c)
Section 8.4.1.1.d stated, The EDGs shall be located in Class 1 structures.
d)
Section I.4.3.14, HVAC Systems, stated: The only HVAC Equipment
required for safe-shutdown are the ECCS Room Coolers, V-AC-5 (Division I)
and V-AC-4 (Division II), located in respective Reactor Building corner rooms
on the 920-foot elevation, and the EDG supply fan, V-SF-9 for EDG No. 12,
and V-SF-10 for EDG No.11.
e)
USAR Table J.4.5-1, Appendix R Safe-shutdown Equipment List, identified
fans V-SF-9 and -10 as safe-shutdown equipment.
The inspectors concluded that the EDG ventilation fans were an auxiliary necessary for
the EDG system and that the term EDGs in the USAR included all of the auxiliaries for
the EDGs; therefore, the ventilation fans were necessary for safe-shutdown of the
reactor (to achieve and maintain cold shutdown).
Based on no actual licensing document specifically mentioning the EDG ventilation,
the necessity of the fans for EDG operability, the USAR references to the fans being
safe-shutdown equipment, and the NSP-1 statement that the full-height wall separated
the EDG systems, the inspectors concluded that the fans and temperature control
dampers should be considered Class 1 equipment and should have been protected from
tornado missiles, as well as the effects of tornadoes on the ventilation ducts as
described in RIS 2006-023, Post Tornado Operability of Ventilating and Air Conditioning
Systems Housed in Emergency Diesel Generator Rooms. However, because this issue
was not clearly defined in the original licensing documents, this will be an unresolved
item pending consultation with NRC headquarters for clarification of whether the EDG
building ventilation was or was not required to be a Class 1 system.
Enclosure
23
.4
Operating Experience
a.
Inspection Scope
The inspectors reviewed six operating experience issues to ensure that 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 2006-22, New Ultra-Low-Sulfur Diesel Fuel Oil Could Adversely Impact Diesel
Engine Performance;
IN 2007-09, Equipment Operability Under Degraded Voltage Conditions;
IN 2008-02, Findings Identified During Component Design Bases Inspections;
IN 2009-02, Biodiesel in Fuel Oil Could Adversely Impact Diesel Engine
Performance;
RIS 2006-023, Post Tornado Operability of Ventilating and Air Conditioning
Systems Housed in Emergency Diesel Generator Rooms; and
SC06-01, Worst Single Failure for Suppression Pool Temperature Analysis.
b.
Findings
No findings of significance were identified.
.5
Modifications
a.
Inspection Scope
The inspectors reviewed three permanent plant modifications related to selected risk-
significant components to verify that the design bases, licensing bases, and performance
capability of the components had not been degraded through modifications. The
modifications listed below were reviewed as part of this inspection effort:
EC 12361, EPU-Provide Operations the Ability to Throttle MO-2020 and
MO-2021;
EC-11734, 1AR Transformer Replacement, Monticello Plant EPU Project; and
SRI-91-010, USAR Update Concerning Diesel Oil Day Tank.
b.
Findings
No findings of significance were identified.
Enclosure
24
.6
Risk-Significant Operator Actions
a.
Inspection Scope
The inspectors performed a margin assessment and detailed review of four risk-
significant, time critical operator actions. These actions were selected from the
licensees PRA rankings of human action importance based on risk achievement worth
values. Where possible, margins were determined by the review of the assumed design
basis and USAR response times and performance times documented by job
performance measures results. For the selected operator actions, the inspectors
performed a detailed review and walk through of associated procedures, including
observing some actions in the plant with an appropriate plant operator to assess
operator knowledge level, adequacy of procedures, and availability of special equipment
where required.
The following operator actions were reviewed:
Loss of Instrument Air;
Operating Suppression Pool Cooling;
Align Diesel Fire Pump in a Station Blackout; and
Time Critical Manual Actions Outside the Main Control Room Before Core
Damage.
b.
Findings
No findings of significance were identified.
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.
b.
Findings
No findings of significance were identified.
Enclosure
25
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 and the control logic changes to allow operations
the ability to manually set an incremental position for the safety-related outboard
containment spray valves.
EC 12361, EPU-Provide Operations the Ability to Throttle MO-2020 and MO-
2021; and
EC 11126, EPU-MOD 11-Balance of Plant Piping Support Modifications.
b.
Findings
(1) Pipe Support Design Deficiencies
Introduction: A finding of very low safety-significance and associated Non-Cited
Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by
the inspectors for the failure to demonstrate pipe supports met their design
requirements. Specifically, the inspectors identified that the calculation for pipe support
SR-526 failed to use the minimum yield strength in determination of the allowable
bending stress of the pipe support SR-526 baseplate as required in the American
Institute of Steel Construction (AISC) code. In addition, the inspectors identified that the
calculation for pipe support PS-16 failed to use the design basis concrete compressive
strength in determination of the anchor bolt allowable as required in the licensees
design specification.
Description: The inspectors identified two examples where pipe supports would not
meet their design requirements. These examples are as follows:
Pipe Support SR-526
The design function of pipe support SR-526 was to hold and maintain SW6-24"-JF
discharge line in position during a seismic Category I design basis event to meet
internal flooding requirements. The failure of this pipe support may create a pipe
break in the SW6-24"-JF piping line, which could cause internal flooding and
subsequent impact on the RHR pumps P-202A/C and core spray pump P-208A.
Also, the failure of pipe support SR-526 due to a seismic Category I design basis
event could impact the RHR heat exchanger E-200A. The inspectors identified a
non-conservative technical error in pipe support SR-526 calculation CA-04-112,
Evaluation of Support SR-526 for Supplementary Load Combination. The
calculation evaluated acceptability of the baseplate connection based on an
allowable bending stress of 0.75 times the actual baseplate material yield stress.
The acceptance criterion for allowable bending stress of a pipe support baseplate
was established in Specification No. MPS-1100, Specification for the Analysis of
Piping and Pipe Support System, Revision 8. Section 8.2.2 of MPS-1100 required
Enclosure
26
the seismic Category I pipe support stress limits to be in accordance with the AISC.
The requirement in the AISC was that the allowable bending stress was 0.75 times
the specified minimum yield stress of the material. The licensee determined that
the calculation CA-04-112 specified the actual material yield stress instead of the
minimum yield stress. The use of actual material yield stress for the evaluation of
the baseplate did not meet AISC code requirements.
The licensee agreed that the use of the Certified Material Test Report or the actual
material yield stress to evaluate the pipe support baseplate was outside the
allowable bending stress acceptance criteria as described in AISC. This issue was
entered in the licensees corrective action program as ARs 01198415 and
01199540. The licensee performed an analysis and determined the pipe support
was operable but nonconforming.
Pipe Support PS-16
The design basis function of pipe support PS-16 was to hold and maintain
PS4-18-ED main steam line in position during a seismic Category I design basis
event. The outboard MSIV AO-2-86D was located between the primary
containment penetration X-7D and pipe support PS-16. The inspectors were
concerned that, if the pipe support did not meet its seismic Category 1 design basis
requirements, the line could be damaged during a design basis event at a location
between the primary containment penetration and the MSIV. This condition could
adversely affect the capability of the outboard MSIV to perform its safety-related
function of providing primary containment isolation.
The inspectors identified a non-conservative technical error in pipe support PS-16
calculation CA-96-147, Evaluation of PS1-18ED, PS2-18ED, PS3-18ED and
PS4-18ED Outside Containment for Revised Turbine Stop Valve Closure Loads.
The calculation evaluated the acceptability of the anchor bolts based on allowables
that use age hardened concrete compressive strength of 6000 pounds per square
inch. The acceptance criteria for anchor bolt allowables were determined from
Specification No. MPS-1100. Section 8.2.7 of MPS-1100 stated the concrete
design strength for determining allowable loads for the anchor bolts shall be 3000
pounds per square inch or 4000 pounds per square inch. The licensee
determined that calculation CA-96-147 specified that use of anchor bolt allowables
based on age hardened concrete compressive strength of 6000 pounds per square
inch instead of 3000 pounds per square inch or 4000 pounds per square inch as
required in Specification No. MPS-1100.
The licensee agreed that the use of age hardened concrete compressive strength
to evaluate pipe support PS-16 anchor bolts was outside the anchor bolt
acceptance criteria established in Specification No. MPS-1100. This issue was
entered in the licensees corrective action program AR 01200215. The licensee
performed an analysis and determined the pipe support was operable but
nonconforming.
Analysis: The inspectors determined that the failure to use minimum specified yield
stress to analyze the applied bending stress on the baseplate for pipe support SR-526
as described in the AISC code and failure to use the design basis concrete compressive
Enclosure
27
strength in determination of the anchor bolt allowable for pipe support PS-16 as required
in MPS-1100 was a performance deficiency.
The performance deficiency for the pipe support SR-526 example was determined to be
more than minor because it was associated with the Mitigating Systems cornerstone
attribute of design control and affected the cornerstone objective of ensuring the
availability, reliability, and capability of the RHR and core spray pumps. Specifically,
compliance with seismic Category I design basis requirements (AISC code
requirements) was to ensure the pipe support SR-526 would function as required during
a seismic Category I design basis event and not cause internal flooding and subsequent
impact on the RHR pumps P-202A/C, and core spray pump P-208A.
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 Systems
cornerstone. The finding screened as of very low safety-significance (Green) because it
was a design deficiency that did not represent an actual loss of safety function. The
inspectors agreed with the licensees position that the pipe support SR-526 was
operable but nonconforming.
The performance deficiency for the pipe support PS-16 example was determined to be
more than minor because it was associated with the Barrier Integrity cornerstone
attribute of design control and affected the cornerstone objective of providing reasonable
assurance that physical design barriers protect the public from radionuclide releases
caused by accidents or events. Specifically, compliance with seismic Category I design
basis requirements was to ensure the pipe support PS-16 would function as required
during a seismic Category I design basis event and not adversely affect the outboard
MSIV AO-2-86D.
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 3b for the Barrier Integrity
cornerstone. The finding screened as of very low safety-significance (Green) because it
was a design deficiency of the physical integrity of the reactor containment that: (1) did
not affect the barrier function of the control room against smoke or a toxic atmosphere;
(2) did not represent an actual open pathway in the physical integrity of reactor
containment; and (3) did not involve an actual reduction in function of hydrogen igniters
in the reactor containment. The inspectors agreed with the licensees position that the
pipe support PS-16 was operable but nonconforming.
Based on the results of the two SDP Phase I screenings, this finding screened as very
low safety-significance (Green). The inspectors determined there was no cross-cutting
aspect associated with this finding because the two examples were legacy design
issues, therefore was not reflective of current performance.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,
in part, that design control measures shall provide for verifying or checking the adequacy
of design, such as by the performance of design reviews, by the use of alternate or
simplified calculational methods, or by the performance of a suitable testing program.
Enclosure
28
Enclosure
29
Contrary to the above, on September 18, 2009 and September 29, 2009, the licensee
failed to demonstrate design adequacy of pipe support SR-526 and pipe support PS-16
was consistent with seismic Category I requirements. Specifically the performance of
design reviews for pipe support SR-526 and pipe support PS-16 were inadequate, in that
design calculation CA-04-112 did not demonstrate that the pipe support SR-526
baseplate would meet AISC code requirements and the design calculation CA-96-147
did not demonstrate the pipe support PS-16 would meet design basis anchor bolt
requirements. Because this violation was of very low safety-significance and it was
entered into the licensees corrective action program (ARs 01198415, 01199540 and
01200215), this violation is being treated as an NCV, consistent with Section VI.A.1 of
the NRC Enforcement Policy. (NCV 05000263/2009007-07)
4OA6 Meeting(s)
.1
Exit Meeting Summary
On October 2, 2009, the inspectors presented the inspection results to Mr. T. OConnor,
and other members of the licensee staff. On December 4, 2009, the inspectors
conducted a re-exit of the inspection results to Mr. T. OConnor, 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.
ATTACHMENT: SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
T. OConnor, Site Vice President
J. Grubb, Plant Manager
P. Albares, Operations Shift Manager
R. Anderson, System Engineering Supervisor
R. Baumer, Regulatory Affairs Compliance Engineer
V. Bhardwaj, System Engineering Manager
N. Haskell, Engineering Director
B. Halvorson, Configuration Management Supervisor
K. Jepson, Business Support Manager
S. Oswald, Regulatory Affairs Analyst
D. Pennington, Mechanical Design Engineer
G. Salamon, Corporate Regulatory Affairs Manager
R. Siepel, Electrical Design Engineer
E. Watzel, Electrical Design Engineering Supervisor
P. Young, Regulatory Programs Supervisor
Nuclear Regulatory Commission
S. Thomas, Senior Resident Inspector, DRP
L. Haeg, Resident Inspector, DRP
Attachment
1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened 05000263/2009007-01
Calculation Errors Associated With the Pneumatic
Pressure Requirements for the Inboard MSIVs05000263/2009007-02
ESW Piping Supports Did Not Meet Seismic
Category 1 Design Basis Requirements05000263/2009007-03
Failure to Adequately Evaluate Minimum Voltage
Available at Safety-Related Electrical Components05000263/2009007-04
Inadequate Testing for MCC Contactors05000263/2009007-05
EDG Fuel Oil Supply System Design does Not Meet
the Single Failure Criteria 05000263/2009007-06
Inadequate Tornado Missile Protection for the EDG
System Components05000263/2009007-07
Pipe Support Design Deficiencies
Closed 05000263/2009007-01
Calculation Errors Associated With the Pneumatic
Pressure Requirements for the Inboard MSIVs05000263/2009007-02
ESW Piping Supports Did Not Meet Seismic
Category 1 Design Basis Requirements05000263/2009007-03
Failure to Adequately Evaluate Minimum Voltage
Available at Safety-Related Electrical Components05000263/2009007-04
Inadequate Testing for MCC Contactors05000263/2009007-07
Pipe Support Design Deficiencies
Discussion
None
Attachment
2
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does
not imply that 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
CA-00-104
Intake Structure Minimum Water Level
0
CA-01-036
Inservice Testing (IST) Pump and Valve Acceptance Criteria
Rounding Evaluation for 4Ih Ten-Year Code Interval
28
CA-01-062
Instrument Uncertainty Calculation
1
CA-01-113
Setpoint for RHR Minimum Flow Switches FS-10-121A, B, C,
and D
13
CA-01-174
Minimum RHRSW Pump Differential Pressure
2
CA-01-177
Determination of Containment Overpressure Required for
Adequate NPSH for Low Pressure ECCS Pumps Updated for
Suction Strainer Debris Loading
8
CA-02-179
125 Volt Div 1 Battery Calculation
1
CA-03-111
EDG Jacket Cooler Maximum Allowed Tubes Plugged
10
CA-03-145
Evaluation of SR-680
1
CA-03-146
Evaluation of New Guides on The High Point Vent
1
CA-03-200
Internal Flooding Evaluation Due To A Postulated Break In 2.5
Fire Line
11
CA-04-010
EOP Calculation Inputs
2
CA-04-103
Emergency Diesel Generator Room Heat-up Air-Cooler Fan-
Flow Analysis
12
CA-04-112
Evaluation of Support SR-526 for Supplementary Load
Combination
0
CA-04-113
Evaluation of Support SR-531 for Supplementary Load
Combination
0
CA-04-114
Evaluation of Support SWH-48A for Supplementary Load
Combination
0
CA-04-127
Determination of Maximum Allowable Outside Air Temperature
13
CA-04-133
AOV Component Calculation, AO-2-86A, B, C, D
15
CA-04-145
EDG Ventilation: Cooling Load and Air Flow Determination
12
CA-04-146
Design of Supports for HVAC Ductwork in DGB (Diesel
Generator Building)
1
CA-06-054
Evaluation of SRP Module J.3 due to Replacement of MO-2009
1
CA-06-081
Evaluation of New Support SR-852 Next to TWH-87
0
CA-06-104
480V MCC to Motor Terminal Voltage Drop
1
CA-07-047
Residual Heat Removal System MOV Performance Analysis
0
CA-08-047
D10 Battery Charger Mounting Evaluation
0
Attachment
3
CALCULATIONS
Number
Description or Title
Revision
CA-08-082
Evaluation of SR-537
0
CA-08-083
Evaluation of SWH-150
0
CA-08-157
Combined AC Model Database
0
CA-09-120
Evaluation of SRP Module F.2 for 120% EPU
0
CA-09-148
Cycle 25 Monticello Specific EOP Calculations, Appendix B.15
0
CA-09-177
Sizing of the Orifice in the SW lines to the RHR Pump Motor
Coolers
0
CA-88-003
RHR Pump Minimum Flow Calculation for IE Bulletin #88-04
Response
0
CA-90-018
Determination of Acceptance Criteria for RHR Pump
Surveillance Testing and Verification of Adequate LPCI Flow
Under Four and Two Pump Operation for SRI No.90-002 and
GE Report NEDC-31786P, Respectively
01/09/91
CA-90-023
Minimum Allowable Fuel Oil Storage Tank Level
2
CA-91-001
125V DC Fault Current
1
CA-91-006
125V DC Battery Charger Sizing
1
CA-91-012
125V DC SBO Load Profile Study
2
CA-91-063
Determination of Uncertainty in RHR Pump D/P Measurement
During Performance of Test 0255-04-111
2
CA-91-086
AC Load Study 1AR No.10 XFMR, LOCA Load 2 Core Spray
Pumps Starting
2
CA-91-092
Plant Fault Study, 2R XFMR, 2RS Reactor In-Line
7
CA-91-093
Plant Fault Study, 1AR Transformer Feed From #10
Transformer
4
CA-91-098
AC Load Study Short Circuit Contribution Database
7
CA-91-100
AC Load Study Transformer Database
7
CA-92-024
125 V DC Battery Sizing Calculation
5
CA-92-036
RHR Pump Room Temperature Analysis
7
CA-92-038
Determination of RHRSW Instrumentation Inaccuracies
4
CA-92-214
RHR System Motor Operated Valve Functional Analysis
11
CA-92-223
Analysis for Mod. 92Q520 Control Circuit Voltage Drop
0
CA-92-224
Emergency Diesel Generator Loading
5
CA-92-250
Project 92Q600 Cable Sizing Calculation
0
CA-92-295
Protection Settings for New LC-103 480V Switchgear Lineup
1
CA-92-299
Stem Thrust Assessment of 16 A/D Globe Valves: MO-2012
and MO-2013
1
CA-93-023
Instrument/Control Setpoint Calculation: Main Air Supply To
Outboard MSIV
1
CA-93-066
AC Loads Study, Degraded Voltage Setpoint, 1R XFMR, LOCA
Load
6
CA-93-084
Hydrogen Generation of No. 11 and No. 12 Battery Rooms
1
Attachment
4
CALCULATIONS
Number
Description or Title
Revision
CA-94-017
Calculation of Alternate Nitrogen System Operability Leakage
Criteria
6
CA-94-037
Calculation of Inboard MSlV Post-LOCA Closing Forces and
Pneumatic Pressure Requirements
5
CA-94-084
Determination of MSIV Stroke Time Acceptance and Setpoint
Bands
7
CA-94-094
MCC Starter Coil Pickup Voltages and Maximum Cable Lengths
0
CA-94-142
Stem Thrust Assessment of 10 A/D Gate Valves: MO-2020,
MO-2021, MO-2022, and MO-2023
2
CA-95-076
Calculation for ASCO Pressure Switch Set Points
0
CA-95-117
Stem Thrust Assessment of 12 A/D Gate Valves: MO-2006
and MO-2007
0
CA-96-012
Minimum Thrust Required to Close Inboard MSIVs and to
Obtain Leak Tight Seat
2
CA-96-015
MO-2006 and MO-2007 Pressure Locking Interim Calc.
0
CA-96-113
Temperature of RHR Rooms During DBA LOCA
0
CA-97-090
AC Voltage Study, 480V Voltage Determination During Diesel
LOOP/LOCA ECCS CS Start Test Conditions
2
CA-97-157
RHR Room Temperature Response to LOCA
2
CA-98-008
Environmental Qualification (50.49) of Automatic Valve
Company Air Control Assembly
5
CA-98-032
Environmental Qualification (50.49) of Namco EA740/EA180
Series Limit Switches
11
CA-98-033
Environmental Qualification (50.49) of Namco EC210 Quick
Disconnects
15
H20-M-001
Cooling Coil Performance at Minimum Water Flow Rate
2
CORRECTIVE ACTION PROGRAM DOCUMENTS
Number
Description or Title
Date
00633248
Supports SWH-42,47 Stanchion do Not Touch Baseplate
02/23/03
00839220
EDG Fuel Oil Transfer System is Vulnerable to Operator Error
06/01/05
01002620
Packing has Slight Leak on MO-2007
11/02/05
01002810
Potential NRC Violation for Sprinkler Obstruction in EDG room
11/03/05
01046881
MO-2007 12 RHR Discharge to Torus has an Oil Leak
08/28/06
01051775
New Fuel Shipment is Ultra-Low-Sulfur-Diesel
09/22/06
01051793
CST Line Above Torus has Wood Wedge Between Wall and
Pipe
09/22/06
01056182
Motor Terminal Voltages Could Drop Below 90% Rated
10/17/06
01061972
Diesel Fuel Oil Pump Disch Pressure Shows Lowering Trend
11/15/06
01103509
Station Transformer Annual Insulating Oil Dielectric Test
07/25/07
01119969
Inconsistencies Between B Manual and TS Bases
12/04/07
Attachment
5
CORRECTIVE ACTION PROGRAM DOCUMENTS
Number
Description or Title
Date
01130761
CV-1729 Not Controlling at 7000 gpm
03/12/08
01135806
Support SWH-42 Stanchion Does Not Touch Baseplate
04/28/08
01135808
Support SWH-47 Stanchion Does Not Touch Baseplate
04/28/08
01146078
High Voltage Spurious Trips-New D10 Charger Bench Test
07/31/08
01149737
Potential Safety Hazard from MO-2013 Insulation Lagging
09/05/08
01151194
Packing Leak on MO-2007, RHR B Outboard Torus Isolation
09/19/08
01152938
11 and 12 EDG-ESW IST Pump Trends
09/30/08
01155238
Station Evaluation Of Industry And Internal OE
10/14/08
01160133
Implementation of Time Critical Operator Actions Ctrl Process
11/21/08
01164976
Unplanned TS Action Entry due to RCIC Inoperability
01/10/09
01170994
Supports SWH-42,47, Inadequate Documentation of Issue
02/26/09
01171900
EC 12361 MO-2020 (DW Spray)-ECN Error
03/05/09
01173431
AO-2-86D Has Excessive Seat Leakage
03/18/09
01175465
OSP-DOL-0543, Rev 2
03/28/09
01186755
NRC Question Re Core Spray Flow Throttling for ASDS
06/24/09
01187408
Documentation of 1AR Transformer Operability Questions
06/29/09
01190108
Station Transformer Annual Insulating Oil Test results
07/20/09
01193840
Error/Wrong Input in Calc 06-104, 480V Motor Term Voltage
08/17/09
01194290
Error in Voltage Drop Calculation CA-06-104
08/24/09
01195281
Cracks in Concrete Pad of 1AR Transformer
08/26/09
01195865
Minimum Voltage for 2R to 1R Auto Transfer
08/31/09
CORRECTIVE ACTION PROGRAM DOCUMENTS GENERATED AS A RESULT OF THE
INSPECTION
Number
Description or Title
Date
01193742
Panel D21 Is Missing Two Bolts
08/15/09
01196091
Minor References List Error in EPU Calc.08-082
09/01/09
01196110
Bus Bracing Ratings Not Documented 4.16 kV Switchgear
and 480V Load Centers
09/02/09
01196224
EPU Calcs Not Listed on ADLs and Stored As Required
09/02/09
01196345
ESW Supports SWH-42 and SWH-47 are Not Functional
09/03/09
01196371
CT Ratio Discrepancy Between Relay Cards and Drawing
09/11/09
01196394
CA-94-037 Calculation Error
09/03/09
01196433
Identified Condition Adverse to Quality Not Corrected
09/03/09
01196448
Discrepancy Between Drawings of MCC 142 for B4208
09/04/09
01196451
EDG Base Tank Volume Calculation CA 90-023
09/03/09
01196495
EDG Fuel Oil XRF Pump Not Included in the EDG Loading
09/04/09
01196513
Stanchion for PI-1982 Only has 4 Bolts Installed
09/04/09
01196531
NE-93194-11 and NE-36438-9 Show Incorrect HP for P-
77/MTR
09/04/09
Attachment
6
CORRECTIVE ACTION PROGRAM DOCUMENTS GENERATED AS A RESULT OF THE
INSPECTION
Number
Description or Title
Date
01196811
Apparent Deficiency in Documentation of EDG Air Start
09/10/09
01196861
CA-00-104, Intake Structure Min Water Level Is Incorrect
09/08/09
01197333
No Documentation for Fuse in D11-22 (Computer Input)
09/17/09
01197339
Inconsistency in Fuse Sizing Div 1 125 VDC D111-04 Circuit
09/11/09
01197431
Undocumented Assumption in Calculation 06-104
09/12/09
01197447
Evaluation of IN 2007-09 Did Not Address Issue
09/03/09
01197792
CA-91-100 2R Impedances Are For 50 MVA, Not 25 MVA
09/15/09
01197923
Mod Did Not Assess Impact to T-44 Tank
09/15/09
01197979
Typo in Ops Man B-09.06-02 for 2R Transformer
09/16/09
01198021
Errors on MCC Schedule Drawings
09/16/09
01198033
CA-07-047 Not Updated with 2003 Test Data Correctly
10/16/09
01198046
D111 Panel Lower Right Side Wing Panel Has Gap
09/16/09
01198056
Incorrect Pressure Term Used in CA-92-214
10/16/09
01198090
Revise CA-92-224 to Remove 2750 kW and 3050 kW
Ratings
09/16/09
01198227
Update CA-03-111 to Incorporate Enhancement
09/17/09
01198266
Weak Link Labeled Incorrectly in MOV Database
09/17/09
01198415
Service Water Line Support May Not Meet Code
Requirements
09/18/09
01198495
CA-94-037 Has Error
09/18/09
01198855
50.59 Did Not Evaluate Use of Age Hardened Concrete
09/21/09
01199213
Ref. for EPU CA-08-081 Documented Incorrectly
09/23/09
01199540
Calc Accept Criteria for Suprt May Not Meet Code Reqmts
09/25/09
01199542
New Crack Not Identified in PUSAR
09/25/09
01199606
Error in 50.59 Screening for EPU EC 11126
09/25/09
01199936
Voltage Drop Evaluation for RHR/RHRSW ASCO Solenoid
Vlvs
09/28/09
01199999
Minor Error in Line Designation Table Drwg ND-57664-1
09/28/09
01200120
Procedure Question Operation of RHRSW at 6000 gpm
09/29/09
01200170
Pickup Voltage on 480V MCC Contactors Not Periodically
Tested
09/29/09
01200215
Incorrect Input was Used in Calculation
09/29/09
01200265
Incorrect Concrete Strength Chosen for 2 EPU Calcs
09/29/09
01200397
Information Lacking on 09-120 to Assess Suitability
09/30/09
01200432
Incorrect Input Used in Calculation No.09-122
09/30/09
01200487
Apparent Conflict Between Calculations
09/30/09
01200539
Revise Procedure 4847-PM (GE 7700 Line MCC) to Require
Contactor Pickup Voltage Test
10/01/09
01200540
Revise Procedure 4027-PM (Klockner-Moeller MCC B34
and B44) to Require Contactor Pickup Voltage Test
10/01/09
Attachment
7
CORRECTIVE ACTION PROGRAM DOCUMENTS GENERATED AS A RESULT OF THE
INSPECTION
Number
Description or Title
Date
01200546
Relays 10A-K70A and10A-K70-B are Agastat E7012AD004
10/01/09
01200682
Support SR-526 is in Contact with Insulation on RHR HX
10/01/09
01200723
Voltage at P-111B Contactor is Less Than 96 Volts
10/02/09
01200725
Instr. Uncertainty Assumed in OPR 01076631 Not Bounding
10/02/09
01202633
B4319 Contactor Failed Degraded Voltage Testing
10/15/09
DRAWINGS
Number
Description or Title
Revision
DNL32274
1AR Transformer Nameplate
E
FSB-0503-1
Fuse Breaker Study
0
FSB-0504-1
Fuse Breaker Study
0
FSB-0507-1
Fuse Breaker Study
1
FSB-0508-1
Fuse Breaker Study
0
M-112
PandID RHR Service Water and Emergency Service Water
Sys
84
M-120
PandID Residual Heat Removal System
76
M-121
PandID Residual Heat Removal System
76
M-133
PandID Diesel Oil System
76
M-134
PandID Fire Protection System Interior Locations
70
M-143
PandID Primary Containment and Atmospheric Control
System
77
M-343
Plans and Sections, HandV Standby Diesel Generator
Building
1/21/69
M-811
PandID Service Water System and Make-Up Intake
Structure
91
NE-100346
Div I and Div II 120V Instrument AC Distribution Panel
Schedules
M
NE-36347-10
No. 142-480V MCC B42
NE-36347-13
No. 133 and No. 143-480V MCC B33 and B43
N
NE-36347-15
No. 134-480V MCC B34
N
NE-36347-1A
480V MCC Schedules
77
NE-36347-8
No. 133-480V MCC B33
79
NE-36394-18
Monticello Nuclear Generating Plant Emergency Service
Water Pumps
76
NE-36404-18
RHR System Auxiliary Controls
L
NE-36404-18A
Schematic Diagram Reactor Auxiliary Systems
K
NE-36771-3
Instrument AC and Uninterruptible AC Panel Schedules
Y10, Y20 andY30
76
NE-93576
Single Line Diagram 480V MCC B34
J
Attachment
8
DRAWINGS
Number
Description or Title
Revision
NF-36176
Generator Aux Transformer and 4160 Volt System Buses 11
and 12
78
NF-36177
Single Line Meter and Relay Diagram 4160 Volt System
Buses No. 13, No. 14, No. 15, and No. 16
77
NF-36298-1
Electrical Load Flow One Line Diagram
81
NF-36298-2
DC Electrical Load Distribution One Line Diagram
78
NL-96763-1
Pipe Support SR-526
A
NL-96792-1
Pipe Support SR-147
A
NX-28921-3
Transformer, Class A, Core form, Outdoor, 3 Phase, 60
Hertz KVA-50000-Namplate Drawing
A
NX-7905-46-11
Elementary Diagram Residual Heat Removal System
J
NX-7905-46-14E
11 RHR Containment Spray Outboard Isolation MO-2020,
Scheme B3339
76
NX-7905-46-16
Residual Heat Removal System Elementary Diagram
J
NX-7905-46-17
Elementary Diagram Residual Heat Removal System
77
NX-7905-46-18
Elementary Diagram Residual Heat Removal System
R
NX-7905-46-19
Elementary Diagram Residual Heat Removal System
P
NX-8714-2-4
MCC B33(A) Turbine Building Elevation 971-0
H
NX-8714-2-5
No. 133-480V MCC B33 (A) Rear Essential Turbine Bldg-
East
78
NX-8714-4-2
Motor Control Center B43-B No. 143 (B) 480V
H
NX-8763-23
V-AC-4 Air Cooling Unit
C
NX-8763-23-1
RHR/Core Spray Pump Room Cooling Coil (V-AC-4)
A
NX-9525-1
RHRSW Pumps - Vertical Pump Dimensions/Layout
76
NX-9525-18
RHRSW Pumps - Bowl Assembly
A
USAR Figure
8.7-1
Instrument AC and Uninterruptible AC Distribution System
Single Line
22
MISCELLANEOUS
Number
Description or Title
Date or
Revision
IST Stroke Time Data for MO2007, MO2013, and MO2020
14 ESW Quarterly Pump Test Data (05/22/07 - 07/22/09)
Phase 2 Risk Evaluation of CAP AR00076349, Loss of
Containment Overpressure Due to Spurious Operation of the
Torus and Drywell Purge and Vent Valves
07/14/09
4 AWI-04.05.12
Replacement of Failed Fuses
4
5828/6453-E-6
Specification for 480 Volt Motor Control Centers for the
Monticello Nuclear Generating Plant - Unit 1
0
8285
Non-Identical Fuse replacement
8
B779R3
60,000 GAL U.G. Oil TK
09/05/67
Attachment
9
MISCELLANEOUS
Number
Description or Title
Date or
Revision
CML
Component Master List for CV-1729 Loop Components: E/P
1729, FT-10-97B, FI-4105, FI-10-132B, DPIC-10-130B, DPI-
10-130B, FI-7188
02/21/05
Outboard MSIV Spring Degradation Evaluation
02/23/09
Evaluation of Inboard MSIV Closure Margin Using Existing
Alternate Nitrogen System Minimum Pressure
09/23/09
EWI-08.15.02
Motor Operated Valve Program Engineering Standards
8
FORM 3090
IST Program Pump Data Sheet for No. 11 RHR Pump (202A)
5
Job No. 5828
Civil Structural Design Criteria for the Monticello Nuclear
Generating Plant-Unit 1
1
LER 2009-01
Containment Overpressure Not Ensured in the Appendix
Analysis
09/18/09
Letter to NRC
IE Bulletin 88-04 Response
07/08/88
Letter to NRC
IE Bulletin 88-04 Final Response
12/13/88
Letter To NSP
Sulzer Bingham Pumps INC., (Vendor Information
Responding to Bulletin 88-04 for RHR Pumps)
11/08/88
MPS-1100
Specification for The Analysis of Piping and Pipe Support
Systems, NPD-M-038
8
MWI-3-M-2.01
AC Electrical Load Study
12
MWI-3-M-2.06
Fuse/Breaker Coordination Study and Electrical Coordination
9
NX-17496-3
MNGP Protective Relay Cards - 4KV
3
NX-17496-4
MNGP Protective Relay Cards - 480V
1
NX-7905-55
11 RHR Pump Curve Vendor Curve (No. 26653)
A
NX-7905-56
11 RHR Pump Curve Vendor Test Data
A
NX-7905-58
13 RHR Pump Curve Vendor Curve (No. 26680)
A
NX-7905-59
Residual Heat Removal Pump 270430
A
NX-7905-70-1
11 RHRSW Pump Curve
03/31/75
NX-7905-72-1
13 RHRSW Pump Curve
03/31/75
SAR01116710
CDBI Focused Self-Assessment
12/17/08
SAR01166267
CDBI FSA Effectiveness and Inspection Readiness
07/21/09
MODIFICATIONS
Number
Description or Title
Date or
Revision
04Q030
EDG Ventilation System Upgrades
0
EPU-MOD 11-Balance of Plant Piping Support Modifications
0
EPU-Provide Operations the Ability to Throttle MO-2020 and
MO-2021
0
Security Improvements - 2009- Force on Force Project
0
EDG Rooms Sprinkler Modification
0
Attachment
10
MODIFICATIONS
Number
Description or Title
Date or
Revision
Modification No. 5 1AR Transformer Replacement,
Monticello Plant EPU Project
4
SRI-91-010
USAR Update Concerning Diesel Oil Day Tank
12/01/92
OPERABILITY EVALUATIONS
Number
Description or Title
Date
LAR Required for Use of TORMIS Code Methodology
02/28/07
CA-03-124
ESW Model ESW2B-Operability Evaluation
06/24/03
CA-03-125
ESW Model ESW1B-Operability Evaluation
06/24/03
FA 010709705
Modify EDG Exhaust Silencer Lines to Restore the EDGs
Within a Reasonable Time Frame
10/01/09
OPR 1076631
Possibility of Inadequate Flow to Room Cooler
03/08/07
OPR 1101934
Operability Determination on 14 RHR Pump
07/22/07
OPR 1130761
Flow Control Problems with B RHRSW Valve CV-1729,
03/20/08
OPR 1169854-1
Outboard MSIVs Have Not Been Tested IAW Testing
Requirements
02/23/09
OPR 1199936-1
Terminal Voltage at SV-1995 and SV-1728
10/02/09
OPR1193840-1
Motors Supplied by 480V MCCs With Voltage Less than 90%
Rated Voltage
08/21/09
PROCEDURES
Number
Description or Title
Revision
0137-15-01
Containment Spray Loop A Isolation Valve Local Leak
Rate Test
10
0137-29
LPCI Loop B Injection Valves Local Leak Rate Test
7
0187-01
11 EDG/ESW Quarterly Pump and Valve Tests
70
0192
Diesel Fuel Quality Check
26
0255-04-IA-1-1
RHR Loop A Quarterly Pump and Valve Tests
78
0255-04-IA-1-2
RHR Loop B Quarterly Pump and Valve Tests
80
1052-03
11 Diesel Generator Auxiliary Systems Test
11
1487
Site Housekeeping Quarterly Inspection
5
4066-PM
D10 Battery Charger Preventive Maintenance
0
4106-01-PM
Emergency Diesel 1 Cycle Maintenance
24
4844-PM
GE Thermal Overload Relay Test Procedure
20
4846-PM
GE/W Molded Case Circuit Breaker Maintenance and Test
Procedure
18
4847-PM
7700 Line Motor Control Center Maintenance Procedure
19
4858-60-PM
1AR Reserve Transformer Maintenance
12
Attachment
11
PROCEDURES
Number
Description or Title
Revision
6-C-3
Diesel Service Oil Pump Tripped
2
6-C-6
Diesel Generator Tank T-45A Level/Flow Low
2
6-C-7
Diesel Generator Tank T-45B Level/Flow Low
3
8196
Temporary Shielding Installation
4
8900
Operation of RCIC without Electric Power
2
8-B-19
ESW Pump 11 Lo Discharge Pressure
4
8-B-20
ESW Pump 11 OL/MAN Override
4
A.6
Acts of Nature
31
B.03.04-06
Residual Heat Removal System - Figures
5
B.08.01.03-01
RHR Service Water System - Function and General
Description
10
B.08.01.03-05
RHR Service Water System-System Operation
39
B.08.04.01-05
Instrument and Service Air
10
B.08.05-05
Fire Protection - System Operation
45
B.08.11-05
Diesel Oil System
17
B.09.06-02
4.16 KV Station Auxiliary
10
B.09.13-06
Instrumentation AC and Uninterruptible AC Distribution
System - Figures
5
C.4-B.08.01.03-
A
Loss of Instrument Air
17
C.4-B.09.02.B
Loss of Normal Offsite Power
11
C.4-C
Shutdown Outside Control Room
32
C.4-H
Restoration Of Plant Loads
13
C.5.1-1000
EOP Introduction
20
C.5.1-1100
RPV Control
8
C.5-3203
Use of Alternate Injection Systems for RPV Makeup
10
C.5-3502
14
FP-E-MOD-04
QF-0515a Design Input Checklist
10
FP-E-MOD-04
QF-0515b Design Input Checklist
3
FP-E-MOD-06
QF-0525 Design Description Form
2
FP-OP-COO-01
Conduct of Operations
6
FP-OP-OL-01
Operability / Functionality Determination
5
Fuel Oil Receiving Quality Check
2
OSP-EDG-0540-
11
11 Emergency Diesel Generator 24 Month Test
2
OWI-03.07
Time Critical Operator Actions
0
Attachment
12
SURVEILLANCES (COMPLETED)
Number
Description or Title
Date or
Revision
0255-05-IA-1-1
A RHRSW Quarterly Pump And Valve Tests
05/15/09
05/27/09
06/11/09
0198-01
125 VDC Battery Capacity Test
03/31/05
4510-PM
Maintenance of Onsite Batteries and Battery Chargers
01/25/07
11 Emergency Diesel Generator 24 Month Test
04/04/09
0187-01B
11 Emergency Diesel Generator/11ESW/Monthly Pump
and Valve Tests
06/14/09
08/09/09
0187-01
11 Emergency Diesel Generator/11ESW/Quarterly
Pump and Valve Tests
07/12/09
WORK ORDERS
Number
Description or Title
Date
00002043
Valcor Solenoid Valve Control Assembly
12/11/01
00134749
Pre-Op Test New 125V DC Division 1 Battery Charger
D10
08/15/08
00200755
PM 4900-1, VOTES for MO2013
05/05/03
00200759
MO2020 VOTES, Inspect LS
05/16/03
00285275-01
PM 4106-01-PM, 11 EDG G-3A, Perform Electrical PM
10/09/06
00293778
P-111A, Rebuild Pump and Return to Stock
06/05/09
00306982
EC210 O Rings On Valcor Control Assembly J-box
03/14/05
00307012
11 RHR Pump Minimum Flow, (PM CV-1994 Air
Operator)
03/14/05
00307804
Make Repairs to Support SWH-47
05/28/03
00307805
Make Repairs to Support SWH-42
05/17/03
00309248
11 RHR Pump Minimum Flow, Baseline Testing on CV-
1994
03/15/05
00311664
D-10, 125VDC Charger for 11 Battery 480V Supply
07/21/04
00311666
MO-1427, RBCCW INL to DW Cooler V-CC-1 480V
Supply
03/29/05
00311667
MO-1429, RBCCW INL to DW Cooler V-CC-3 480V
Supply
03/29/05
00311670
P-88B, ECCS Area Drain Pump B (480V Supply)
05/26/04
00336541
Monitor ADS Pneumatic Supply
01/22/08
00343696-01
PM 4030-01-PM, 11 Diesel Generator Control Panel C-
91/93
03/07/09
00343716
CV-1994, Valve Position Indication Test, (0255-03-IA-
2B)
05/27/09
00343738
CV-1728 Diagnostic Test: Post Outage PM work.
04/04/09
00343827-01
PM 4851-12 (52-304) MCC-133A Feeder Breaker
Cubicle
03/09/09
Attachment
13
Attachment
14
SURVEILLANCES (COMPLETED)
Number
Description or Title
Date or
Revision
00343889
Limit Switch (AO-2-86D)
04/23/09
00343994
PM 4058-6 (B RHR RM Air Cooling Unit V-AC-4)
07/24/09
00344785
Monitor ADS Pneumatic Supply
03/18/08
00349292
Monitor ADS Pneumatic Supply
05/14/08
00352042
Perform Coupling Inspection of P-77
07/15/08
00352697
Perform Coupling Inspection of P-11
07/08/08
00354013
D10 125 VDC Charger 24 Month Capacity Test
06/13/08
00354414
MO2007 Rebuild Actuator and Perform 4900-01-PM
04/04/09
00355172
CV-1728, Perform IandCPM 7070 (A)on RSW-1
Instruments
09/15/08
00356497
CV-1729, Comprehensive 11 RHRSW Pump and Valve
Tests
09/10/08
00356504
11 EDG Auxiliary Systems Test
03/29/09
00356593-01
PM 4109-01-PM, 11 EDG 12 Year Maintenance
03/07/09
00356627
Battery 11 Modified Performance Test
03/29/09
00359064-11
Perform Acceptance Testing of 1AR Transformer
07/24/09
00362759
025504-III-3A Comprehensive 13 RHR Pump and
Valve Test
12/09/08
00362760
CV-1729, Comprehensive 11 RHRSW Pump and Valve
Tests
12/26/08
00363012
Division 1 LPCI Pump Discharge Flow-Low Bypass
Channel Calibration, (ISP-RHR-0547-01)
12/15/08
00363156
0255-04-IA-11 RHR Loop A QRTRLY Pump and Valve
Test
07/20/09
00366681
CV-1729, Comprehensive 11 RHRSW Pump and Valve
Tests
03/13/09
00374280
11 and 12 125 VDC Battery Weekly Surveillance
06/30/09
00375976
11 EDG/ESW Quarterly Pump and Valve Tests
07/10/09
00377539
11 EDG/ESW Quarterly Pump and Valve Tests
08/09/09
00377780-01
Perform Major PM on Spare Breaker LCB-037
04/04/09
00379154
0255-04-IA-11 RHR Loop A QRTRLY Pump and Valve
Test
12/09/08
00385117
Diesel Fuel Oil Receiving Quality Check
05/19/09
00403241
11 RHR Pump Minimum Flow, (CV-1994 failed to
Operate)
09/17/04
09906286
Measure Air Flow on V-AC-4 and V-AC-5
07/01/09
LIST OF ACRONYMS USED
Agencywide Document Access Management System
American Institute of Steel Construction
Action Request
American Society of Mechanical Engineers
Component Design Bases Inspection
CFR
Code of Federal Regulations
Control Power Transformer
Division of Reactor Projects
Division of Reactor Safety
EC
Engineering Change
Emergency Service Water
GL
Generic Letter
IEEE
Institute of Electrical and Electronic Engineers
IN
Information Notice
IMC
Inspection Manual Chapter
IR
Inspection Report
Inservice Test
kV
Kilovolt
Load Center
Loss of Coolant Accident
Loss of Off-site Power
Motor Control Center
Motor-Operated Valve
Non-Cited Violation
National Electrical Manufactures Association
Net Positive Suction Head
NRC
U.S. Nuclear Regulatory Commission
Publicly Available Records
Preventative Maintenance
psig
Pounds Per Square Inch Gauge
Residual Heat Removal Service Water
Regulatory Information Issue
Station Blackout
Significance Determination Process
Safety Evaluation Report
Standardized Plant Analysis Risk
Thermal Overload
TS
Technical Specification
Updated Safety Analysis Report
Unresolved Item
Vac
Volts Alternating Current
Vdc
Volts Direct Current
Attachment
15
T. OConnor
-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 Agencywide
Documents Access and Management System (ADAMS), accessible from the NRC Web site 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 No. 50-263
License No. DPR-22
Enclosure:
Inspection Report 05000263/2009007
w/Attachment: Supplemental Information
cc w/encl:
Distribution via ListServ
DISTRIBUTION:
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ROPreports Resource
DOCUMENT NAME: G:\\DRS\\Work in Progress\\MON 2009-007 CDBI DRS IR AXD.doc
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OFFICE
RIII
RIII
RIII
NAME
ADunlop:ls
DPassehl
AMStone
DATE
12/9/09
12/10/09
01/06/10
OFFICIAL RECORD COPY